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Sample records for oecd-csni loca standard

  1. Lessons learned from OECD/CSNI ISP on small break LOCA: final report

    International Nuclear Information System (INIS)

    1996-07-01

    This document presents an overview of the results obtained from five recent OECD/CSNI International Standard Problems (ISPs) dealing with phenomenologies typical of Small Break LOCA in PWR nuclear power plants of western design. The experiment in four Integral test Facilities, Lobi, Spes, Bethsy and Lstf and the recorded data from a steam generator tube rupture transient in the Belgian PWR of Doel, were taken as reference for the calculations. Relevant hardware characteristics of the facilities and of the plant are firstly given, including the correlation between key thermalhydraulic phenomena and the reference experimental scenarios. A statistical evaluation of the general data connected with each ISP is then presented. The lessons learned from the ISPs are then considered. Four areas have been identified: code deficiencies and capabilities, scaling of the data, progress in code capabilities and various additional aspects

  2. Comparison report on OECD-CSNI containment standard problem N.1

    International Nuclear Information System (INIS)

    Winkler, W.

    1980-05-01

    The technical purpose of this containment standard problem N.1 was to compare experimental results of history of pressure, temperature, pressure difference and water mass after a steam line rupture within a chain of six subsequent compartments (simplified integral test) with the corresponding results of best-estimate post-test calculations from computer codes for three different time intervals. 11 countries took part in the comparison, using 11 different computer codes and several versions

  3. Comparison report of the OECD/CSNI international standard problem 21 (Piper-one experiment PO-SB-7). Volume 1 comparison report. Volume 2 evaluation of code accuracy in the prediction of ISP 21

    International Nuclear Information System (INIS)

    1989-11-01

    The present report deals with the comparison of 6 blind predictions, submitted by 5 participants, and the experimental results measured during the test PO-SB-7 performed in PIPER-ONE facility. The PIPER-ONE apparatus is an experimental simulator of a General Electric BWR. The test PO-SB-7 simulates a SB-LOCA originated by a break in one recirculation line of the reference BWR-6 plant, without intervention of high pressure ECCS. The overall activity constitutes the CSNI ISP-21. The main parts of the report are: a) outline of the test facility and of the PO-SB-7 experiment; b) overview of input models used by participants; c) evaluation of participant predictions on the basis of one-by-one comparison with selected experimental trends; d) evaluation of present code capabilities and accuracy, on the basis of the overall comparison between measured data and participants double blind predictions. Finally, a judgement is given in relation to the overall value of the activity

  4. Fuel safety criteria and review by OECD / CSNI task force

    International Nuclear Information System (INIS)

    Van Doesburg, W.

    1999-01-01

    Full text of publication follows: with the advent of advanced fuel and core designs, and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a general feeling that safety margins have been or are being reduced. Historically, fuel safety margins were defined by adding conservatism to the safety limits, which in turn were also fixed in a conservative manner, here, the expression 'conservatism' expresses the fact that bounding or limiting numbers were chosen for model parameters, plant and fuel design data, and fuel operating history values. Unfortunately, as these conservatisms were not quantified (or quantifiable), the amount of safety available or the reduction thereof is difficult to substantiate. For the regulator, it is important to know the margin available with the utilities' request for approval of new fuel or methods; likewise, for the utility and vendor it is important to know what margins exist and what they are based on, to identify in which direction they can make further progress and optimize fuel and fuel cycle cost. Naturally, each party involved will have to decide on how much margin should be in place, to establish operational criteria and ensure that these can actually be met during operation. To assess the margins issue, safety criteria themselves need to be reviewed first. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel available at that time - mostly at zero exposure. Of course, verification was performed as designs progressed in later years, primarily with the aim to be able to prove that safety criteria were adequate as long as the said conservatisms would be retained, and not with the aim to reestablish limits. The mandate to the OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) is to assess the adequacy of existing fuel safety criteria, in view of the 'new design' elements (new

  5. Proceedings of the OECD/CSNI specialists meeting on fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru; Yamano, Norihiro; Sugimoto, Jun [eds.

    1998-01-01

    The OECD/CSNI Specialists Meeting on Fuel Coolant Interactions (FCI) was held at Tokai-mura in Japan on May 19 through 21, 1997, and attended by 80 participants from 14 countries and one international organizations. In the meeting 36 papers were presented followed by active discussions in six sessions on various aspects of FCI issues, such as reactor application, premixing, propagation/trigger, experiments and code/models. At the end of the Meeting, the participants have reached to the consensus on the summary and recommendations, which consists of the following items; (1) We find no new evidence that would change or violate the conclusion of SERG-2 (1996) that alpha-mode failure is not risk significant. (2) Significant progress has been made since the Santa Barbara meeting (1993). (3) Several areas have been identified, which need further investigations to understand the basic FCI phenomena, and to improve the modeling. (4) We recommend maximizing open communication between various research groups in order to accelerate the resolution of the remaining issues. (5) We recommend that the next specialist meeting be held within 3 to 5 years in order to synthesize the activities described above. (J.P.N.)

  6. Analysis of OECD/CSNI ISP-42 phase A PANDA experiment using coupled code R5G (RELAP5-GOTHIC)

    International Nuclear Information System (INIS)

    Bencik, V.; Debrecin, N.; Grgic, D.; Bajs, T.

    2010-01-01

    In the paper, the results of the analysis of OECD/CSNI ISP-42 Phase A experiment at PANDA facility using stand-alone codes RELAP5/mod3.3 and GOTHIC 7.2b as well as coupled code R5G (RELAP5/mod3.3-GOTHIC 7.2b) are presented. PANDA is a large-scale thermal-hydraulic test facility installed at PSI (Paul Scherrer Institute) in Switzerland. The OECD/CSNI ISP-42 test consists of six sequential phases (Phase A through F). The present work deals with the post-test calculation of the Phase A, including the break of the main steam line and the Passive Containment Cooling (PCC) System Start-Up. The objective of the test is to investigate the start-up phenomenology of passive cooling system when steam is injected into cold vessel filled with air. The calculation was performed using stand-alone RELAP5/mod3.3 and GOTHIC 7.2b models, and then the same calculation was performed using coupled code with RELAP5 being responsible for reactor part of the model and GOTHIC being responsible for containment part of the model. The prediction capability, running time and modeling aspects were discussed for all three cases. (authors)

  7. Fuel safety criteria technical review - Results of OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria

    International Nuclear Information System (INIS)

    Hollasky, N.; Valtonen, K.; Hache, G.; Gross, H.; Bakker, K.; Recio, M.; Bart, G.; Zimmermann, M.; Van Doesburg, W.; Killeen, J.; Meyer, R.O.; Speis, T.

    2000-01-01

    With the advent of advanced fuel and core designs, the adoption of more aggressive operational modes and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a concern if safety margins have remained adequate. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel that was available at that time, mostly with unirradiated specimens. Verification was of course performed as designs progressed in later years, however mostly with the aim to be able to prove that these designs adequately complied with existing criteria, and not to establish new limits. The OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) was therefore given the mandate to technically review the existing fuel safety criteria, focusing on the 'new design' elements (new fuel and core design, cladding materials, manufacturing processes, high burnup, MOX, etc.) introduced by the industry. It should also identify if additional efforts may be required (experimental, analytical) to ensure that the basis for fuel safety criteria is adequate to address the relevant safety issues. In this report, fuel-related criteria are discussed without attempting to categorize them according to event type or risk significance. For each of these 20 criteria, we present a brief description of the criterion as it is used in several applications along with the rationale for having such a criterion. New design elements, such as different cladding materials, higher burnup, and the use of MOX fuels, can affect fuel-related margins and, in some cases, the criteria themselves. Some of the more important effects are mentioned in order to indicate whether the criteria need to be re-evaluated. The discussion may not cover all possible effects, but should be sufficient to identify those criteria that need to be addressed. A summary of these discussions is given in Section 7. As part

  8. Post-LOCA long term cooling performance in Korean standard nuclear power plants

    International Nuclear Information System (INIS)

    Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

    1999-01-01

    The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break LOCA and large break LOCA. The RELAP5/MOD3.2.2 code is used to calculate the LTC sequences based on the LTC plan of the KSNPP. A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important manual action including the safety injection tank isolation in LTC procedure is investigated

  9. Application of the Relap5-3D to phase 1 and 3 of the OECD-CSNI/NSC PWR MSLB benchmark related to TMI-1

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.; Spadoni, A.; Hassan, Y.

    2001-01-01

    The Relap5-3D, the latest in the series of the Relap5 code, distinguishes from the previous versions by the fully integrated, multi-dimensional thermalhydraulic and kinetic modeling capability. It has been applied to Phase I and III of OECD-CSNI/ NSC PWR MSLB Benchmark adopting the same thermalhydraulic input deck already used with Relap5/Parcs and Relap5/Quabbox coupled codes during the previous MSLB analysis. The OECD jointly with the US NRC proposed the PWR MSLB Benchmark in order to gather a common understanding about the coupling between thermal hydraulics and neutronics, and evaluating the behavior of this transient with different coupled codes, giving emphasis to the 3-D modeling. This paper deals with the application of Relap5-3D code to phase I and III of the PWR MSLB Benchmark. The Relap5-3D is a thermal hydraulics-neutronics internally coupled code, the thermal hydraulics module is the INEEL version of Relap and the neutronics module is derived from NESTLE multi-dimension kinetics code. (author)

  10. Evaluation of post-LOCA long term cooling performance in Korean standard nuclear power plants

    International Nuclear Information System (INIS)

    Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

    2001-01-01

    The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT) isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences. (author)

  11. Post-test sensitivity analysis of OECD/CSNI ISP42 panda experiment by Relap5 code

    International Nuclear Information System (INIS)

    Zanocco, P.; D'Auria, F.; Galassi, G.M.

    2001-01-01

    The present document deals with Relap5/Mod3.2 analysis of the International Standard Problem (ISP-42) exercise performed in PANDA facility on April 21-22, 1998. PANDA is installed at PSI (Paul Scherrer Institute). PANDA is a large-scale thermal-hydraulic test facility suitable for the simulation of passive containment for Advanced Light Water Reactors (ALWR). The work focuses phase A of the ISP-42 experiment, including the break in the main steam line, and the Passive Containment Cooling System Start-Up. The objective is to investigate the start-up phenomenology of passive cooling system when steam is injected into cold vessel filled with air and to observe the resulting system behavior. A detailed nodalization was set-up at the University of Pisa, in order to model 3-D flow paths with a 1-D code. The comparison between pre-test predictions and experimental data is discussed. Overall time behavior is reasonably well predicted, showing a rather good and robust overall code behavior in the simulation of the global test scenario. The results of a preliminary post-test analysis are discussed, including the comparison with the experimental data. (authors)

  12. Mitigating Capability Analysis during LOCA for Korean Standard Nuclear Power Plants in Containment Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Park, Soo Yong; Kim, D. H.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    The objective of this paper is to establish Containment spray operational technical bases for the typical Korean Standard Nuclear Power plants (Ulchin units 3 and 4) by modeling the plant, and analyzing a loss of coolant accident (LOCA) using the MAAP code. The severe accident phenomena at nuclear power plants have large uncertainties. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand severe accident sequences and to assess the accident progression accurately by computer codes. Furthermore, it is important to attain the capability to analyze a advanced nuclear reactor design for a severe accident prevention and mitigation.

  13. Simulation of international standard problem no. 44 open tests using Melcor computer code

    International Nuclear Information System (INIS)

    Song, Y.M.; Cho, S.W.

    2001-01-01

    MELCOR 1.8.4 code has been employed to simulate the KAEVER test series of K123/K148/K186/K188 that were proposed as open experiments of International Standard Problem No.44 by OECD-CSNI. The main purpose of this study is to evaluate the accuracy of the MELCOR aerosol model which calculates the aerosol distribution and settlement in a containment. For this, thermal hydraulic conditions are simulated first for the whole test period and then the behavior of hygroscopic CsOH/CsI and unsoluble Ag aerosols, which are predominant activity carriers in a release into the containment, is compared between the experimental results and the code predictions. The calculation results of vessel atmospheric concentration show a good simulation for dry aerosol but show large difference for wet aerosol due to a data mismatch in vessel humidity and the hygroscopicity. (authors)

  14. ISP - 26 OECD/NEA/CSNI International standard problem n. 26. Rosa-4 LSTF Cold-Leg Small-Break Loca experiment. Comparison report

    International Nuclear Information System (INIS)

    1992-02-01

    This report is the final comparison report for the CSNI ISP-26 which was performed for a 5 pc cold-leg small-break LOCA experiment conducted in the ROSA-IV Large scale test facility (LSTF), and simulation of the thermal-hydraulic response of a pressurized water reactor during a small break loss-of-coolant accident or an operational transient. The facility has an overall scaling factor of 1/48, with hot and cold legs sized to conserve the volume scaling. Both the initial steady-state conditions and the test procedures are designed to minimize the effects of scaling compromises. 10 seconds into the break, the turbine throttle valve was closed at a pressurizer pressure of 12.97 MPa. The turbine bypass was inactive, since loss-of-offsite power was assumed concurrently with the scram. The reactor coolant pumps stopped at 265 seconds. The core was uncovered between 120 seconds and 155 seconds into the break. The comparison report presents all the nineteen submitted calculations. A summary description of the participants as well as the computer codes and input decks used by them is provided

  15. Analysis of factors affecting the LOCA test quality

    International Nuclear Information System (INIS)

    Wang Lu

    2008-01-01

    Localization of nuclear safety-related equipment has become an important way of nuclear power development in China. To meet this demand, the competence should be promoted in the following two areas, one is to develop the capability of R and D and manufacturing of nuclear safety-related equipment, the other is to implement equipment qualification according to relevant codes and standards. As LOCA test is one of the most important parts in the qualification test of nuclear safety-related equipment, the main factors related with the quality of the LOCA test are analyzed in this paper, and this may be a reference to improve the skills in designing, constructing and operating LOCA test devices. (authors)

  16. In-core LOCA (PTR) analysis with poisoned moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, T. M.; Choi, J. H.; Kim, Yun Ho; Choi, Hoon

    2005-01-01

    CANDU reactors have been analyzed and evaluated for the postulated in-core LOCA while the reactor is operating normally with low moderator poison concentration. However, when the reactor is operating with relatively large amounts of boron and/or gadolinium poisons in the moderator, the assessment for fuel integrity was required for pressure tube rupture (PTR) accident. The methodology of in-core LOCA analysis with poisoned moderator is developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for CANDU reactor recently. The developed methodology and results are presented

  17. Progress in realistic LOCA analysis

    Energy Technology Data Exchange (ETDEWEB)

    Young, M Y; Bajorek, S M; Ohkawa, K [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    1994-12-31

    While LOCA is a complex transient to simulate, the state of art in thermal hydraulics has advanced sufficiently to allow its realistic prediction and application of advanced methods to actual reactor design as demonstrated by methodology described in this paper 6 refs, 5 figs, 3 tabs

  18. OECD/CSNI Workshop on Best Estimate Methods and Uncertainty Evaluations - Workshop Proceedings

    International Nuclear Information System (INIS)

    2013-01-01

    Best-Estimate Methods plus Uncertainty Evaluation are gaining increased interest in the licensing process. On the other hand, lessons learnt from the BEMUSE (NEA/CSNI/R(2011)3) and SM2A (NEA/CSNI/R(2011)3) benchmarks, progress of UAM benchmark, and answers to the WGAMA questionnaire on the Use of Best-Estimate Methodologies show that improvements of the present methods are necessary and new applications appear. The objective of this workshop is to provide a forum for a wide range of experts to exchange information in the area of best estimate analysis and uncertainty evaluation methods and address issues drawn-up from BEMUSE, UAM and SM2A activities. Both, improvement of existing methods and recent new developments are included. As a result of the workshop development, a set of recommendations, including lines for future activities were proposed. The organisation of the Workshop was divided into three parts: Opening session including key notes from OECD and IAEA representatives, Technical sessions, and a Wrap-up session. All sessions included a debate with participation from the audience constituted by 71 attendees. The workshop consisted of four technical sessions: a) Development achievements of BEPU methods and State of the Art: The objective of this session was to present the different approaches to deal with Best Estimate codes and uncertainties evaluations. A total of six papers were presented. One initial paper summarized the existing methods; the following open papers were focused on specific methods stressing their bases, peculiarities and advantages. As a result of the session a picture of the current State of the Art was obtained. b) International comparative activities: This session reviewed the set of international activities around the subject of BEPU methods benchmarking and development. From each of the activities a description of the objectives, development, main results, conclusions and recommendations (in case it is finalized) was presented. This session comprised a total of four papers. c) Applications: licensing, safety analysis support, regulatory body views and industry activities: This session put the focus on the application of current methods in safety analyses. Contribution from industry, technical safety organizations and regulatory bodies was present. As a result, a view of the penetration of BEPU methods in current safety analyses was obtained, as well as an indication of the evolution in the near future. Elements such as licensing practices, assessment process, etc. were considered. This session comprised nine papers. d) BEPU methods extension to new fields: Since the early days of BEPU methods development it was realized that this approach could be extended and its benefits applied to other areas beyond the classic deterministic environment. This session was focused on the extension of BEPU methods beyond this point. The session developed the extension to CFD, quantification of global safety margins, use of PSA, etc. with a total of seven papers

  19. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    International Nuclear Information System (INIS)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts' meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes

  20. OECD/CSNI specialist meeting on nuclear aerosols in reactor safety - Summary and Conclusions

    International Nuclear Information System (INIS)

    Allelein, Hans-Josef; Boulaud, Denis; Guentay, Salih; Dehbi, Abdelouahab; Hontanon, Esther; Jokiniemi, Jorma; Jones, Alan V.; Koroll, Grant W.; Tinkler, Charles G.; Schaperow, J.; Royen, Jacques

    1999-01-01

    The Third OECD Specialist Meeting on Nuclear Aerosols in Reactor Safety was organised in Cologne, Germany, from 15-18 June 1998, in collaboration with the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH. It was attended by sixty-five specialists representing thirteen OECD Member countries and the Commission of the European Communities. Thirty-nine papers were presented, in eight sessions. The meeting was concluded by a general discussion devoted to the following topics: - What has been solved up to the level of plant applications and accident management? - Where is more work needed for plant applications? Over the last eight years significant progress has been made in source term modelling and code development. Results have been consolidated in codes which are being used for plant calculations to address safety issues. For example, two countries used their source term codes last year to assess the impact of heating from deposited fission products on the potential for steam generator tube rupture. Also, another source term code was used to assess the need of sprays for fission product removal in the proposed design of a new type of reactor. Over the next few years, experiments and model development will continue, with more emphasis on application to risk-important severe accident scenarios

  1. OECD/CSNI Workshop on In-Vessel Core Debris Retention and Coolability - Summary and Conclusions

    International Nuclear Information System (INIS)

    Behbahani, Ali-Reza; Drozd, Andrzej; Kim, Sang-Baik; Micaelli, Jean-Claude; Okkonen, Timo; Sugimoto, Jun; Trambauer, Klaus; Tuomisto, Harri

    1999-01-01

    In the spring of 1994 an OECD Workshop on Large Pool Heat transfer was held in Grenoble. The scope of this workshop was the investigation of (1) molten pool heat transfer, (2) heat transfer to the surrounding water, and (3) the feasibility of in-vessel core debris cooling through external cooling of the vessel. Since this time, experimental test series have been completed (e.g., COPO, ULPU, CORVIS) and new experimental programs (e.g., BALI, SONATA, RASPLAV, debris and gap heat transfer) have been established to consolidate and expand the data base for further model development and to improve the understanding of in-vessel debris retention and coolability in a nuclear power plant. Discussions within the CSNI's PWG-2 and the Task Group on Degraded Core Cooling (TG-DCC) have led to the conclusion that the time was ripe for organizing a new international Workshop with the objectives: - to review the results of experimental research that has been conducted in this area; - to exchange information on the results of member countries experiments and model development on in-vessel core debris retention and coolability; - to discuss areas where additional experimental research is needed in order to provide an adequate data base for analytical model development for core debris retention and coolability. The scope of this workshop was limited to the phenomena connected to in-vessel core debris retention and coolability and did not include steam explosion and fission product issues. The workshop was structured into the following sessions: Key note papers; Experiments and model development; Debris bed heat transfer; Corium properties, molten pool convection and crust formation; Gap formation and gap cooling; Creep behaviour of reactor pressure vessel lower head; Ex-vessel boiling and critical heat flux phenomena; Scaling to reactor severe accident conditions and reactor applications. Compared to the previous workshop held in Grenoble in 1994, large progress has been made in the understanding and the modelling of several phenomena involved in the domain of interest. They concern: the corium properties, the molten pool convection, the gap formation and cooling, the creep behaviour of the lower head, the ex-vessel critical heat flux, and they are discussed in the session specific conclusions (chapter 2.2). Some conclusions have been reached and some outstanding questions have been identified during the general discussion : Plant specific studies performed on core melt retention indicate with high confidence level that the in-vessel retention concept is possible for low (∼600 MWel) power reactors. Convective corium pool behaviour with all chemical and physical aspects requires still confirmatory research. Such as being carried out in OECD Rasplav project. The pool formation and the initial conditions in the lower plenum are plant and accident sequence dependent and are subject to large uncertainties. Since in-vessel coolability cannot be ensured (e.g., for high power plants), it is highly desirable to understand RPV lower head mechanical behaviour (i.e., size, location and time to failure) and to improve the modelling of creep rupture behaviour. For severe accident analyses different approaches are useful: Obtain physical understanding by well designed experiments and detailed code calculations; Apply integral codes to obtain scenario and system effects such as timing; Complement remaining shortcomings by separate engineering calculations when necessary. One of the most important work is to develop a consistent severe accident management strategy for each individual plant. This ensures the above discussed chemical and physical phenomena do not cause confusion among the reactor safety society

  2. Proceedings of the OECD/CSNI workshop on transient thermal-hydraulic and neutronic codes requirements

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, D.

    1997-07-01

    This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.

  3. Proceedings of the OECD/CSNI specialist meeting on advanced instrumentation and measurement techniques

    Energy Technology Data Exchange (ETDEWEB)

    Lehner, J [comp.

    1998-09-01

    In the last few years, tremendous advances in the local instrumentation technology for two-phase flow have been accomplished by the applications of new sensor techniques, optical or beam methods and electronic technology. The detailed measurements gave new insight to the true nature of local mechanisms of interfacial transfer between phases, interfacial structure and two-phase flow turbulent transfers. These new developments indicate that more accurate and reliable two-phase flow models can be obtained, if focused experiments are designed and performed by utilizing this advanced instrumentation. The purpose of this Specialist Meeting on Advanced Instrumentation and Measurement Techniques was to review the recent instrumentation developments and the relation between thermal-hydraulic codes and instrumentation capabilities. Four specific objectives were identified for this meeting: bring together international experts on instrumentation, experiments, and modeling; review recent developments in multiphase flow instrumentation; discuss the relation between modeling needs and instrumentation capabilities, and discuss future directions for instrumentation development, modeling, and experiments.

  4. Proceedings of the OECD/CSNI specialist meeting on advanced instrumentation and measurement techniques

    International Nuclear Information System (INIS)

    Lehner, J.

    1998-09-01

    In the last few years, tremendous advances in the local instrumentation technology for two-phase flow have been accomplished by the applications of new sensor techniques, optical or beam methods and electronic technology. The detailed measurements gave new insight to the true nature of local mechanisms of interfacial transfer between phases, interfacial structure and two-phase flow turbulent transfers. These new developments indicate that more accurate and reliable two-phase flow models can be obtained, if focused experiments are designed and performed by utilizing this advanced instrumentation. The purpose of this Specialist Meeting on Advanced Instrumentation and Measurement Techniques was to review the recent instrumentation developments and the relation between thermal-hydraulic codes and instrumentation capabilities. Four specific objectives were identified for this meeting: bring together international experts on instrumentation, experiments, and modeling; review recent developments in multiphase flow instrumentation; discuss the relation between modeling needs and instrumentation capabilities, and discuss future directions for instrumentation development, modeling, and experiments

  5. Proceedings of the OECD/CSNI specialists meeting on boron dilution reactivity transients

    International Nuclear Information System (INIS)

    1997-06-01

    The purpose of the meeting was to bring together experts involved in the different activities related to boron dilution transients. The experts came from all involved parties, including research organizations, regulatory authorities, vendors and utilities. Information was openly shared and discussed on the experimental results, plant and systems analysis, numerical analysis of mixing and probability and consequences of these transients. Regulatory background and licensing implications were also included to provide the proper frame work for the technical discussion. Each of these areas corresponded to a separate session. The meeting focused on the thermal-hydraulic aspects because of the current interest in that subject and the significant amount of new technical information being generated

  6. OECD/CSNI specialist meeting on advanced instrumentation and measurements techniques: summary and conclusions

    International Nuclear Information System (INIS)

    1997-01-01

    This specialist meeting on Advanced Instrumentation and Measurements Techniques was held in Santa Barbara (USA) in 1997 and attracted some 70 participants in ten technical sessions and a session of the round table discussions, with a total of 41 papers. It was intended to bring together the international experts in multi-phase flow instrumentation, experiment and modeling to review the state-of-the-art of the two-phase flow instrumentation methods and to discuss the relation between modeling needs and instrumentation capabilities. The following topics were included: Modeling needs and future direction for improved constitutive relations, interfacial area transport equation, and multi-dimensional two-fluid model formulation; local instrumentation developments for void fraction, interfacial area, phase velocities, turbulence, entrainment, particle size, thermal non-equilibrium, shear stress, nucleation, condensation and boiling; global instrumentation developments for void fraction, mass flow, two-phase level, non-condensable concentration, flow regimes, low flow and break flow; relation between modeling needs and instrumentation capabilities, future directions for experiments focused on modeling needs and for instrumentation developments

  7. OECD/CSNI specialist meeting on fuel coolant interactions: summary and conclusions

    International Nuclear Information System (INIS)

    1997-01-01

    Research activities and interest on fuel-coolant interaction (FCI) have been increased and broadened since the last CSNI Specialist Meeting held in January 1993. Significant experimental and analytical research has been performed in many OECD countries and others. The growing international interest is, in large part, due to the emphasis on broader aspects of FCI ranging from melt quenching and coolability to energetic explosions (both in- and ex-vessel), and their relevance and applications to next-generation reactor design as well as accident management strategies. The objectives of the meeting are to review the knowledge and to obtain consensus on the phenomenology of FCI and in predicting FCI behavior in LWRs severe accidents; to identify those areas of FCI phenomena and prediction which are important for reactor safety but still poorly understood and require further study with clear methodologies; to inform the community and the regulatory agencies of the status of FCI issues, especially in the application to accident management and future reactor designs. The various sessions are: reactor applications, pre-mixing, propagation / trigger, experiments

  8. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  9. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    International Nuclear Information System (INIS)

    Cho, Jaehyun; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-01-01

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  10. Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr; Park, Jin Hee; Kim, Dong-San; Lim, Ho-Gon

    2017-04-15

    Highlights: • We quantified the LOCA core damage frequency based on the best-estimated success criteria analysis. • The thermal-hydraulic analysis using MARS code has been applied to Korea Standard Nuclear Power Plants. • Five new event trees with new break size boundaries and new success criteria were developed. • The core damage frequency is 5.80E−07 (/y), which is 12% less than the conventional PSA event trees. - Abstract: A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4 in and by only a high pressure safety injection in LOCAs of 0.8–9.4 in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety

  11. Progress in realistic LOCA analysis

    International Nuclear Information System (INIS)

    Young, M.Y.; Bajorek, S.M.; Ohkawa, K.

    2004-01-01

    In 1988 the USNRC revised the ECCS rule contained in Appendix K and Section 50.46 of 10 CFR Part 50, which governs the analysis of the Loss Of Coolant Accident (LOCA). The revised regulation allows the use of realistic computer models to calculate the loss of coolant accident. In addition, the new regulation allows the use of high probability estimates of peak cladding temperature (PCT), rather than upper bound estimates. Prior to this modification, the regulations were a prescriptive set of rules which defined what assumptions must be made about the plant initial conditions and how various physical processes should be modeled. The resulting analyses were highly conservative in their prediction of the performance of the ECCS, and placed tight constraints on core power distributions, ECCS set points and functional requirements, and surveillance and testing. These restrictions, if relaxed, will allow for additional economy, flexibility, and in some cases, improved reliability and safety as well. For example, additional economy and operating flexibility can be achieved by implementing several available core and fuel rod designs to increase fuel discharge burnup and reduce neutron flux on the reactor vessel. The benefits of application of best estimate methods to LOCA analyses have typically been associated with reductions in fuel costs, resulting from optimized fuel designs, or increased revenue from power upratings. Fuel cost savings are relatively easy to quantify, and have been estimated at several millions of dollars per cycle for an individual plant. Best estimate methods are also likely to contribute significantly to reductions in O and M costs, although these reductions are more difficult to quantify. Examples of O and M cost reductions are: 1) Delaying equipment replacement. With best estimate methods, LOCA is no longer a factor in limiting power levels for plants with high tube plugging levels or degraded safety injection systems. If other requirements for

  12. CANDU fuel behaviour under LOCA conditions

    International Nuclear Information System (INIS)

    Kohn, E.

    1989-07-01

    This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

  13. Parameter study on the influence of prepressurization on LWR fuel rod behaviour during normal operation and hypothetical LOCA

    International Nuclear Information System (INIS)

    Fuchs, H.P.; Brzoska, B.; Depisch, F.; Sauermann, W.

    1978-01-01

    To analyse the influence of prepressurization on fuel rod behaviour, a parametric study has been performed considering the effects of the as-fabricated fuel rod internal prepressure on the normal operation and postulated LOCA red behaviour of a 1300 MWe1 KWU standard nuclear power plant pressurized water reactor. A reduction of prepressurization in the analysed range results in a negligible worsened normal operation behaviour whereas the LOCA behaviour is improved significantly. (author)

  14. LOCA and RIA studies at JAERI

    International Nuclear Information System (INIS)

    Sugiyama, Tomoyuki; Nagase, Fumihisa; Nakamura, Jinichi; Fuketa, Toyoshi

    2004-01-01

    To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA) is being studied at the Japan Atomic Energy Research Institute (JAERI). The LOCA program consists of integral thermal shock tests and other separate tests for oxidation rate and mechanical property of fuel claddings. Prior to the tests on irradiated claddings, the tests have been conducted on non-irradiated claddings to examine separate effects of corrosion and hydrogen absorption during reactor operation. The tests on irradiated claddings have recently been started and results have been obtained. As for an RIA study, a series of experiments with high burnup fuel rods is being performed by using pulse irradiation capability of the NSRR. This paper presents recent results obtained from the LOCA and RIA studies at JAERI. (Author)

  15. Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10. Final report

    International Nuclear Information System (INIS)

    Fields, M.B.; Kudrick, J.A.

    1984-08-01

    This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases and conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff approved acceptance criteria for LOCA-related hydrodynamic loads are provided in an appendix

  16. Mark III LOCA-related hydrodynamic load definition. Generic technical activity B-10

    International Nuclear Information System (INIS)

    1984-02-01

    This report, prepared by the staff of the Office of Nuclear Reactor Regulation and its consultants at the Brookhaven National Laboratory, provides a discussion of LOCA-related suppression pool hydrodynamic loads in boiling water reactor (BWR) facilities with the Mark III pressure-suppression containment design. Its issuance completes NRC Generic Technical Activity B-10, Behavior of BWR Mark III Containment. On the basis of certain large-scale tests conducted between 1973 and 1979, the General Electric Company developed LOCA-related hydrodynamic load definitions for use in the design of the standard Mark III containment. The staff and its consultants have reviewed these load definitions and their bases conclude that, with a few specified changes, the proposed load definitions provide conservative loading conditions. The staff-approved acceptance criteria for LOCA-related hydrodynamic loads are provided in Appendix C of this report

  17. NPP Krsko small break LOCA analysis

    International Nuclear Information System (INIS)

    Mavko, B.; Petelin, S.; Peterlin, G.

    1987-01-01

    Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

  18. Prediction of LOCA Break Size Using CFNN

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Kim, Dong Yeong; Na, Man Gyun [Chosun University Gwangju (Korea, Republic of)

    2016-05-15

    The NPPs have the emergency core cooling system (ECCS) such as a safety injection system. The ECCS may not function properly in case of the small break size due to a slight change of pressure in the pipe. If the coolant is not supplied by ECCS, the reactor core will melt. Therefore, the meltdown of reactor core have to be prevented by appropriate accident management through the prediction of LOCA break size in advance. This study presents the prediction of LOCA break size using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model is a data-based method that requires data for its development and verification. The data were obtained by numerically simulating severe accident scenarios of the optimized power reactor (OPR1000) using MAAP code, because real severe accident data cannot be obtained from actual NPP accidents. The CFNN model has been designed to rapidly predict the LOCA break size in LOCA situations. The CFNN model was trained by using the training data set and checked by using test data set. These data sets were obtained using MAAP code for OPR1000 reactor. The performance results of the CFNN model show that the RMS error decreases as the stage number of the CFNN model increases. In addition, the performance result of the CFNN model presents that the RMS error level is below 4%.

  19. Replacement divider plate performance under LOCA loading

    International Nuclear Information System (INIS)

    Huynk, H.M.; MClellan, G.H.; Schneider, W.G.

    1997-01-01

    A primary divider plate in a nuclear steam generator is required to perform its partitioning function with a minimum of cross leakage, without degradation in operating performance and without loss of structural integrity resulting from normal and accident loading. The design of the replacement divider plate for normal operating conditions is discussed in some detail in reference 1 and 2. This paper describes the structural response of the replacement divider plate to the severe loading resulting from a burst primary pipe. The loads for which the divider plate structural performance must be evaluated are mild to severe differential pressure transients resulting from several postulated sizes and types of pipe break scenarios. In the unlikely event of a severe Loss of Coolant Accident (LOCA) the divider plate or parts thereof must not exit the steam generator nor completely block the outlet nozzle. For the milder LOCA loads, the integrity of the divider plate and seat bars must be maintained. Analysis for the milder LOCA loads was carried out employing a conservative approach which ignores the actual interaction between the structure and the primary fluid. For these load cases it was shown that the divider plate does not become disengaged from the seat bars. For the more severe pipe breaks, the thermal-hydraulic analysis was coupled iteratively with the structural analysis, thereby taking into account divider plate deformation, in order to obtain a better prediction of the behaviour of the divider plate. In this manner substantial reduction in divider plate response to the more severe LOCA loading was achieved. It has been shown that, for the case of a postulated large LOCA (100% reactor inlet header), the disengagement of the divider plate from the seat bars resulted in an opening smaller than 1% of the divider plate area. (author)

  20. LOCA analysis of the IRIS reactor

    International Nuclear Information System (INIS)

    Bajs, T.; Grgic, D.; Cavlina, N.

    2003-01-01

    The IRIS reactor (International Reactor Innovative and Secure) is an integral, light water cooled, medium power reactor. IRIS has been selected as an International Near Term Deployable (INTD) reactor, within the Generation IV International Forum activities. The IRIS concept addresses the key-requirements defined by the US DOE for next generation reactors, i.e. enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). An innovative safety approach has been developed to mitigate the IRIS response to small-to-medium Loss of Coolant Accident (LOCA). This strategy is based on the interaction of IRIS compact containment with the reactor vessel to limit initial blowdown, and on depressurization through the use of a passive Emergency Heat Removal System (EHRS). A small Automatic Depressurization System (ADS) provides supplementary depressurization capability. A pressure suppression system is provided to limit the pressure peak following the initial blowdown to well below the containment design limit. The ultimate result is that during a small-to-medium LOCA, the core remains covered for an extended period of time, without credit for emergency water injection or external core makeup. The IRIS LOCA response is based on 'maintaining water inventory' rather than on the principle of safety injection. This novel safety approach poses significant issues for computational and analysis methods since the IRIS vessel and containment are strongly coupled, and the system response is based on the interaction between the two. The small break LOCA was calculated using RELAP5/mod3.3 and GOTHIC codes. Break of the largest line connected to the IRIS Reactor Pressure Vessel (RPV) was analyzed. The results of the calculations confirmed good performance of the IRIS system during LOCA. (author)

  1. RELAP 5 Simulations of a hypothetical LOCA in Ringhals 2

    International Nuclear Information System (INIS)

    Caraher, D.

    1987-01-01

    RELAP5 simulations of a hypothetical LOCA in Ringhals 2 were conducted in order to determine the sensitivity of the calculated peak cladding temperature (PCT) to Appendix K requirements. The PCT was most sensitive to the assumed model decay heat: Changing from the 1979 ANS Standard to 1.2 times the 1973 Standard increased the PCT by 70 to 100K. After decay heat, the two parameters which affected the PCT the most were steam generator heat transfer and heat transfer lockout. The PCT was not sensitive to the assumed pump rotor condition (locked vs coasting); nor was it sensitive to a modest amount (5 to 10%) of steam generator tube plugging. (author)

  2. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

  3. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Suzuki, H.; Hatamiya, S.; Murase, M.

    1988-01-01

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

  4. Design basis neutronics calculations for NRU-LOCA experiments

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Jenquin, U.P.; McNair, G.W.; Perry, R.T.; Trapp, T.J.; Zimmerman, M.G.

    1979-08-01

    The report describes the neutronics analysis for the LOCA simulation experiments in the NRU reactor. The experimental program will provide greater understanding of nuclear fuel assembly behavior during the heatup, reflood and quench sequence of a hypothetical LOCA. The decay heat and stored heat, which are the energy source in a LOCA will be simulated by fission heat provided by the NRU reactor. The reactor, the test and test operation are described

  5. Critical heat flux concerns during the flow instability phase of a DEGB LOCA

    International Nuclear Information System (INIS)

    Shadday, M.A. Jr.

    1990-08-01

    Arguments are presented that support the proposal that a separate burnout risk analysis, for the Flow Instability (FI) phase of a LOCA, not be required for reactor restart. With expected reactor power limits, flow instability will occur before critical heat flux (CHF). Since FI power limits preclude the occurrence of flow instability in a bounding accident, a DEGB LOCA, the risk of CHF and attendant burnout is negligible. A review of RDAP data revealed that in the past reactor assemblies operated at flow and power conditions similar to those expected in a LOCA without burnout occurring. This is strong bounding empirical evidence, without the scaling concerns of laboratory experiments. A bounding analysis of the influences of assembly non-idealities on CHF, power tilts, and channel eccentricity, is included. The margin between operating heat fluxes, during the postulated LOCA, and CHF was quantified by scoping calculations. Based on measured azimuthal power variations, the local heat flux would have to be more than 20 standard deviations above the calculated mean heat flux for CHF to occur

  6. Estimation of LOCA break size using cascaded Fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    Operators of nuclear power plants may not be equipped with sufficient information during a loss-of-coolant accident (LOCA), which can be fatal, or they may not have sufficient time to analyze the information they do have, even if this information is adequate. It is not easy to predict the progression of LOCAs in nuclear power plants. Therefore, accurate information on the LOCA break position and size should be provided to efficiently manage the accident. In this paper, the LOCA break size is predicted using a cascaded fuzzy neural network (CFNN) model. The input data of the CFNN model are the time-integrated values of each measurement signal for an initial short-time interval after a reactor scram. The training of the CFNN model is accomplished by a hybrid method combined with a genetic algorithm and a least squares method. As a result, LOCA break size is estimated exactly by the proposed CFNN model.

  7. Primary LOCA in VVER-1000 by pressurizer PORV failure

    International Nuclear Information System (INIS)

    Sabotinov, L.; Lutsanych, S.; Kadenko, I.

    2013-01-01

    The paper presents the calculations and analysis of the design basis accident of a standard VVER-1000/V320 reactor with inadvertent opening and stuck in open position of the pressurizer pilot operated relief valve (PORV). The objective is the independent assessment of this accident applying the French best estimate thermal-hydraulic computer code CATHARE 2 and verification to meet the safety criteria for such kind of the accident. The 'Inadvertent opening and stuck in open position of PORV' is a design basis accident classified as Medium Break Loss of Coolant Accident (MB LOCA) with the equivalent diameter of the break D- 68 mm. This accident is particularly interesting to be calculated and analyzed, because it took place at operating NPP with VVER-1000 reactors (Rovno NPP) in 2009. The calculations have been carried out with conservative conditions as usual for DBA analysis. The NPP model corresponds to a real VVER-1000/V320 configuration and comprises all safety systems, adopted for one of the latest CATHARE 2 versions. The results of CATHARE 2 calculations are compared with available results of RELAP5 calculations. There is similarity of the thermal-hydraulic parameters behavior, but also some differences can be observed basically due to the break flow prediction, which causes differences in primary pressure evaluation. Both calculations show that there is no boiling crisis in the reactor core and reliable cooldown is achieved. The calculations performed with CATHARE2 code demonstrate the ability of the code to predict reasonably the break flow, pressures, temperatures etc. for considered LOCA scenario and to be applied for safety studies

  8. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  9. Proceedings of the joint CSNI/CNRA workshop on redefining the large break LOCA: technical basis and its implications

    International Nuclear Information System (INIS)

    2003-01-01

    viewpoints and experiences of the largest utility. The presentation summarized the experi ence data for known degradation mechanisms. The EDF presentation also covered leak before-break concept which is one of the integral elements (along with the in-service inspections and leak detection) in redefining the LB LOCA. The SKI presentation discussed the experience with the degradation specific in-service inspection programs. The presentation also covered a number of efforts that are currently underway to develop risk-informed ISI program. It should be recognized that any risk informed approach to in-service inspection programs must be based on an extensive experience, which at present will probably not cover every possible degradation mechanism. The Framatome presentation described the standardize criteria, methods, and procedures for assuring the design and operational adequacy of reactor coolant pressure boundary leak detection systems used in plants. The presentation also covered a new system based on the measurements of local humidity. 3. What are possible new definitions for the LB-LOCA? What are their implications for current and future reactors? Two papers were presented which highlighted different approaches to incorporating a change in the LB-LOCA definition into a plant. The first paper, 'Slovak approach during the gradual upgrading of Bohunice V 1', by Mr Kliment, described a programme of back fits to an operating plant, to increase the DBA LB-LOCA from 32 mm equivalent to 200 mm. The second paper, 'Westinghouse Owner's Group Large Break Loca Redefinition Program', presented by Mr Bastien, described the simplifications of design and operation that would be achieved if the DBA LB-LOCA were (significantly) reduced in size

  10. Mechanical interaction between fuel pins and assemblies during LOCA in BWR

    International Nuclear Information System (INIS)

    Jonsson, T.

    1978-10-01

    The size of the rod elongation by oxidation is so large that deformation of a standard BWR fuel element with tie rods in the outer row will surely occur during a LOCA transient typical for BWRs with external pumps. Available data does not however show whether this deformation will occur early in the transient or during the cooling. Combined effects of thermal expansion of zircaloy and expansion due to oxidation and dissolution of oxygen can be expected to be large enough to cause rod bowing early in a LOCA transient. It is however not impossible that observed residual expansion of zircaloy tubes to a dominating extent are caused through expansion of zirconium oxide during cool-down. Length measurements of zircaloy tubes during a transient are desirable. (author)

  11. Effect of air on speed of insulating material deterioration under simulated LOCA environment. [Gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Kusama, Yasuo; Yagi, Toshiaki; Ito, Masayuki; Okada, Sohei; Yoshikawa, Masato (Japan Atomic Energy Research Inst., Takasaki, Gunma. Takasaki Radiation Chemistry Research Establishment)

    1982-12-01

    To examine the quality approval testing method for the electric cables used for nuclear reactors, various covering insulating materials employed for the cables have been investigated from all angles. The factors which are considered to affect the deterioration of cable materials in a simulated LOCA (loss of coolant accident) environmental test are numerous. This paper reports on the result of investigation on the effect of air on the rate of deterioration of various organic materials usually used as the insulating and covering materials for the cables. Five kinds of polymer sheets (1 mm thick) used for reactor cables were employed as samples. The samples of both standard compounding ratio and the compounding ratio for practical reactor use were tested. As the deterioration prior to LOCA simulation, the thermal deterioration corresponding to 40 years aging (at 121 deg C for 7 days) was given, and subsequently, 50 Mrad gamma -irradiation at 1 Mrad/h was performed in the air. After that, the samples were subject to LOCA simulated environment. Since the results were different according to the kinds of samples, those are described separately for Hypalon, ethylene propylene rubber, cross-linked polyethylene, chloroprene and silicone rubber. The existence of air under LOCA environment accelerated the deterioration of insulation materials except silicone rubber, though its influence differed to the polymers. These materials swelled in the presence of air, and the degree of swelling increased with the temperature, having the close relation to oxidation deterioration. Polyethylene was more susceptible to the effect of air, and silicone rubber was rather stable. The samples of fire-retardant compounding ratio more swelled by water absorption than those of standard compounding ratio.

  12. Influence of LOCA simulating conditions on the variation of electrical characteristics of insulating materials

    International Nuclear Information System (INIS)

    Okada, Sohei; Yoshikawa, Masato; Ito, Masayuki; Kusama, Yasuo; Yagi, Toshiaki

    1982-01-01

    The authors have examined the variation of insulation resistance when the sheets of insulating materials and cables were exposed to various LOCA simulating environment. This report describes the summarized results obtained so far for ethylene propylene rubber (EPR) which is important as an insulating material of cables. The samples used were an EPR sheet of standard compound ratio, 2 kinds of EPR sheets of practical compound ratio, 6 types of PH cables (fire-retardant, EPR insulated, chlorosulphonated polyethylene sheathed cable) produced for trial as reactor use, and 6 kinds of EPR sheets of the same composition as the cable core. To discuss the difference of insulation resistance change, the logarithmic mean of the ratio of 1 min values to initial insulation resistance rho/rhosub(o) was used. PWR LOCA-simulating environment was used, while the thermal aging in the air at 121 deg C for 7 days and 50 Mrad irradiation in the air at room temperature were given as the predeterioration. The effect of LOCA-simulation period in the simultaneous method without air, in which steam and radiation were given in parallel, the difference in the experimental results of cables and sheets, the effect of air, the comparison of the simultaneous method with the sequential method in which LOCA-simulating steam was applied after the irradiation in the air and the reverse sequential method (dielectric property measurements) are described. Under the existence of air, the sequential method seems to be a good simulation condition for the simultaneous method, though many experiments are required further. (Wakatsuki, Y.)

  13. Influence of LOCA simulating conditions on the variation of electrical characteristics of insulating materials

    Energy Technology Data Exchange (ETDEWEB)

    Okada, Sohei; Yoshikawa, Masato; Ito, Masayuki; Kusama, Yasuo; Yagi, Toshiaki

    1982-12-01

    The authors have examined the variation of insulation resistance when the sheets of insulating materials and cables were exposed to various LOCA simulating environment. This report describes the summarized results obtained so far for ethylene propylene rubber (EPR) which is important as an insulating material of cables. The samples used were an EPR sheet of standard compound ratio, 2 kinds of EPR sheets of practical compound ratio, 6 types of PH cables (fire-retardant, EPR insulated, chlorosulphonated polyethylene sheathed cable) produced for trial as reactor use, and 6 kinds of EPR sheets of the same composition as the cable core. To discuss the difference of insulation resistance change, the logarithmic mean of the ratio of 1 min values to initial insulation resistance rho/rhosub(o) was used. PWR LOCA-simulating environment was used, while the thermal aging in the air at 121 deg C for 7 days and 50 Mrad irradiation in the air at room temperature were given as the predeterioration. The effect of LOCA-simulation period in the simultaneous method without air, in which steam and radiation were given in parallel, the difference in the experimental results of cables and sheets, the effect of air, the comparison of the simultaneous method with the sequential method in which LOCA-simulating steam was applied after the irradiation in the air and the reverse sequential method (dielectric property measurements) are described. Under the existence of air, the sequential method seems to be a good simulation condition for the simultaneous method, though many experiments are required further.

  14. Small break LOCA analysis for RCP trip strategy for YGN 3 and 4 emergency procedure guidelines

    International Nuclear Information System (INIS)

    Suh, Jong Tae; Bae, Kyoo Hwan

    1995-01-01

    A continued operation of RCPs during a certain small break LOCA may increase unnecessary inventory loss from the RCS causing a severe core uncovery which might lead to a fuel failure. After TMI-2 accident, the CEOG developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2) in response to NRC requests and incorporated it in the generic EPG for CE plants. The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis determines the RCP trip setpoint and demonstrates the safe operational aspects of RCP trip strategy during a small break LOCA for YGN 3 and 4. The trip setpoint of the first two RCPs for YGN 3 and 4 is calculated to be 1775 psia in pressurizer pressure based on the limiting small break LOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 3 and 4 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at worst time. Also, the YGN 3 and 4 RCP trip strategy demonstrates that both the 10 CFR 50.46 requirements on PCT and the ANSI standards 58.8 requirements on operator action time can be satisfied with enough margin. Therefore, it is concluded that the T2/L2 RCP trip strategy with a trip setpoint of 1775 psia for YGN 3 and 4 can provide improved operator guidance for the RCP operation during accidents. 11 figs., 4 tabs., 9 refs. (Author)

  15. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1984-01-01

    Hydrogen generation during a PWR LOCA has been estimated for design basis accident and for two more severe hypothetical accidents. Hydrogen production during design basis accident is a rather slow mechanism, allowing in the worst case, 15 days to connect a hydrogen recombining unit to the containment atmosphere monitoring system. Hydrogen generated by steam oxidation during more severe hypothetical accidents was found limited by steam availability and fuel melting phenomena. Uncertainty is, however, still remaining on corium-zirconium-steam interaction. In the worst case, calculations lead to the production of 500 kg of hydrogen, thus leading to a volume concentration of 15% in containment atmosphere, assuming homogeneous hydrogen distribution within the reactor building. This concentration is within flammability limits but not within detonation limits. However, hydrogen detonation due to local hydrogen accumulation cannot be discarded. A major uncertainty subsisting on hydrogen hazard is hydrogen distribution during the first hours of the accident. This point determines the effects and consequences of local detonation or deflagration which could possibly be harmful to safeguard systems, or induce missile generation in the reactor building. As electrical supply failures are identified as an important contributor to severe accident risk, corrective actions have been taken in France to improve their reliability, including the installation of a gas turbine on each site to supplement the existing sources. These actions are thus contributing to hydrogen hazard reduction

  16. Safety studies on LOCA for N.S. Mutsu

    International Nuclear Information System (INIS)

    Kawasaki, Masayuki; Yaguchi, Shinnosuke

    1978-01-01

    A number of safety studies are under way concerning the reactor plant of N.S. Mutsu. One such study relates to Loss of Coolant Accidents (LOCA), which has been conducted to cover mainly the two subjects of experiments to ascertain the integrity of stainless steel fuel cladding under the action of the Emergency Core Cooling System (ECCS), and analysis of containment integrity following a LOCA. The stainless steel cladding tests were conducted to test swelling, rupture, oxidation and compression characteristics. Few reports are known to have been published in this domain, so that the present results should prove useful for future studies related to ECCS evaluation analyses on stainless steel fuel cladding. The containment integrity analysis covered variations of containment pressure and temperature following a LOCA, performed separately for short- and long-term periods. Estimates were also made on the changes in the hydrogen concentration present inside the containment after a LOCA. The results obtained should serve in determining the characteristic response to LOCA of marine reactor plants

  17. Preliminary Results Of LOCA Problem For APR1400 Reactor

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Le Thi Thu; Vo Thi Huong; Le Van Hong

    2011-01-01

    Several features of NPP with APR1400 nuclear reactor during a loss of coolant accident (LOCA) are investigated in this study. The report describes some main design characteristics of an engineering safety systems of APR1400 and the thermal hydraulic calculation results for steady-state using MARS and RELAP/SCDAPSIM codes. Large Break LOCA accident has been analyzed and evaluated based on acceptable criteria for ECCS given by US NRC. The results from cold leg break LOCA with broken area of 0.0465 m 2 in case of high pressure safety injection system (HPSI) failed to operate or 2 and 4 HPSI pumps are activated. The preliminary results of this work is a part of collaboration between INST researchers and KAERI experts in using RELAP tool for safety analysis of NPPs. (author)

  18. Audit calculation for the LOCA methodology for KSNP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Un Chul; Park, Chang Hwan; Choi, Yong Won; Yoo, Jun Soo [Seoul National Univ., Seoul (Korea, Republic of)

    2006-11-15

    The objective of this research is to perform the audit regulatory calculation for the LOCA methodology for KSNP. For LBLOCA calculation, several uncertainty variables and new ranges of those are added to those of previous KINS-REM to improve the applicability of KINS-REM for KSNP LOCA. And those results are applied to LBLOCA audit calculation by statistical method. For SBLOCA calculation, after selecting BATHSY9.1.b, which is not used by KHNP, the results of RELAP5/Mod3.3 and RELAP5/MOD3.3ef-sEM for KSNP SBLOCA are compared to evaluate the conservativeness or applicability of RELAP5/MOD3.3ef-sEM code for KSNP SBLOCA. The result of this research can be used to support the activities of KINS for reviewing the LOCA methodology for KSNP proposed by KHNP.

  19. An IPSN research programme to resolve pending LOCA issues

    International Nuclear Information System (INIS)

    Mailliat, A.; Grandjean, C.; Clement, B.

    2001-01-01

    Studies performed in IPSN and elsewhere pointed out that high burnup may induce specific effects under LOCA conditions, especially those related with fuel relocation. Uncertainties exist regarding how much these effects might affect the late evolution of the accident transient and the associated safety issues. IPSN estimates that a better knowledge of specific phenomena is required in order to resolve the pending uncertainties related to LOCA criteria. IPSN is preparing the so called APRP-Irradie (High Burnup fuel LOCA) programme. One of the important aspect of this programme is in-pile experiments involving bundle geometries in the PHEBUS facility located at Cadarache, France. A feasibility study for such an experimental programme is underway and should provide soon, a finalized project including cost and schedule aspects. (authors)

  20. Uncertainties in radioactivity release from LWR plants under LOCA conditions - magnitude and consequences

    International Nuclear Information System (INIS)

    Mattila, L.J.

    1977-01-01

    Standardized, deterministic, and supposedly conservative calculation methods and parameter values are applied in radiological safety analyses required for licensing individual nuclear power plants. As realistic as possible and comprehensive analyses are, however, absolutely necessary for many purposes, such as developing improved designs, comparisons between nuclear and non-nuclear power plant alternatives or entire energy production strategies, and also formulating improved acceptance criteria for plant licensing. A specific type of LOCA, called design basis accident (DBA), has obtained an exceptionally important status in the licensing procedure of light water reactor nuclear power plants. This postulated accident has a decisive influence on plant siting and on the design of the various engineered safety features. To avoid certain potential negative effects of the highly standardized guideline-based accident analysis procedure - such as introduction of apparent design ''improvements'', wrong priorization of research efforts, etc. - and to provide a realistic view about the safety of light water reactors to supplement the conservative results from regulatory analyses, a comprehensive understanding of the radiological consequences of LOCA's is indispensable. Estimates of fission product release from LWR plants under different LOCA conditions are associated with uncertainties due to deficient knowledge and truly random variability. The following steps of the fission product transport chain are discussed: generation of activity, fission product release from fuel to fuel pin voids prior to the accident, fuel rod puncturing and fission product release from punctured rods during the accident, further release from fuel during the transient, transport to the containment and finally removal in and leakage from the containment. Numerical examples are given by comparing assumptions, parameter values, and results from the following four analyses: the present guideline

  1. MELCOR ex-vessel LOCA simulations for ITER+

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

    1995-01-01

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

  2. BWR 1 % main recirculation line break LOCA tests, RUNs 917 and 918, without HPCS at ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Okazaki, Motoaki; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1988-07-01

    In a case of small break loss-of-coolant accident (LOCA) at a boiling water reactor (BWR) system, it is important to lower the system pressure to cool down the reactor system by using either the high pressure core spray (HPCS) or the automatic depressurization system (ADS). The report presents characteristic test results of RUNs 918 and 917, which were performed at the rig-of-safety assessment (ROSA)-III program simulating a 1 % break BWR LOCA with an assumption of HPCS failure, and clarifies effects of the ADS delay time on a small break LOCA. The ROSA-III test facility simulates principal components of a BWR/6 system with volumetric scaling factor of 1/424. It is experimentally concluded that the ADS delay time shorter than 4 minutes results in a similar PCT as that in a standard case, in which the PCT is observed after actuation of the low pressure core spray (LPCS). And the ADS delay time longer than 4 minutes results in higher PCT than in the standard case. In the latter, the PCT depends on the ADS time, a 220 K higher PCT, for example, in a case of 10 minutes ADS delay compared with the standard case. (author) 52 refs. 299 figs

  3. Studies in Phebus reactor of fuel behaviour upon LOCA conditions

    International Nuclear Information System (INIS)

    Manin, A.; Del Negro, R.; Reocreux, M.

    1980-09-01

    The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

  4. A LOCA analysis for AHWR caused by ECCS header rupture

    International Nuclear Information System (INIS)

    Chatterjee, B.; Gawai, Amol; Gupta, S.K.; Kushwaha, H.S.

    2000-01-01

    Loss of coolant accident (LOCA) analyses for the proposed 750 MWth Advanced Heavy Water Reactor (AHWR), initiated by the rupture of 8 inch NB ECCS header has been carried out. This paper narrates the description of AHWR and associated ECCS, postulated scenario with which the analyses is carried out, results, discussion and conclusion

  5. Practical illustration of the traditional vers. alternative LOCA embrittlement criteria

    International Nuclear Information System (INIS)

    Vrtilkova, V.; Novotny, L.; Hamouz, V.; Doucha, R.; Tinka, I.; Macek, J.; Lahovsky, F.

    2005-01-01

    Evaluation of LOCA time behaviour is usually based on traditional embrittlement criterion, represented by the equivalent cladding reacted (ECR) limit 17 % (18 %) at the peak cladding temperature below 1204 0 C (1200 0 C). From different existing correlations for evaluation the ECR, the correlations of Baker-Just, Cathcart and VNIINM (Bibilashvili) are discussed here. Results, obtained by these correlations, are illustrated for typical and atypical LOCA courses analysed for the WWER 440 plant. An approach to assess these correlations from the viewpoint of violation of the observed criterion is presented. This approach is based on determination of the temperature vers. time of exposition, when the criterion limit is reached. Reasons leading to necessity of alternative criterion proposal are summarised. This criterion for LOCA events evaluation, including corresponding correlation, is proposed on the basis of the long-term experimental research of cladding materials at UJP Praha. The computational results, obtained according to this alternative criterion, are illustrated for the same courses of LOCA events as for traditional criteria and traditional correlations. Proposed criterion is also confronted with the other discussed criteria in accordance with mentioned approach presented in this paper. The characteristic experimental results and key findings are summarised. They substantiate and support the proposed alternative criterion. An advantage of the criterion is its independence on ECR, on hydrogen and oxygen content and on oxidation history, and its applicability to current Zr-based alloy cladding materials as well. This applicability is kept while preserving the simplicity of the criterion using. (author)

  6. Operator reliability analysis during NPP small break LOCA

    International Nuclear Information System (INIS)

    Zhang Jiong; Chen Shenglin

    1990-01-01

    To assess the human factor characteristic of a NPP main control room (MCR) design, the MCR operator reliability during a small break LOCA is analyzed, and some approaches for improving the MCR operator reliability are proposed based on the analyzing results

  7. Leakage rate from LOCA-aged inflatable airlock seals Pickering NGS 'B' personnel doors

    International Nuclear Information System (INIS)

    Fayle, G.W.; Cordingley, D.C.

    1985-01-01

    In order to demonstrate to the Atomic Energy Control Board that an air-lock inflatable seal will function after a LOCA exposure, an inflatable seal intended for personnel doors at the Pickering NGS 'B' was exposed to the thermal/moisture conditions of the LOCA requirement. While attending to determine the post-LOCA leakage rate it was found that additional leaks developed during each post-LOCA inflation/deflation cycle. The seal had been significantly and irreparably deteriorated by the LOCA exposure. The test has demonstrated that this type of LOCA exposed seal should not be expected to withstand either additional pressure above 207 kPa or additional inflation/deflation cycling. A higher inflation pressure and/or cycling will reduce the likelihood of a post-LOCA seal retaining an inflation pressure sufficient to prevent leakage across the seal

  8. A study on the effect of the CHF correlations to the LOCA analysis

    International Nuclear Information System (INIS)

    Kim, Ho Kee

    1998-02-01

    The critical heat flux (CHF) is a major parameter which determines the cooling performance and therefore the prediction of CHF is of importance for the design and safety analysis in boiling systems; such as nuclear reactors, conventional boilers, and other various two-phase flow systems. Until now, many CHF correlations have been developed and for the actual design a correlation has been selected in consideration of its characteristics. For the analysis of Loss of Coolant Accident (LOCA) in a Nuclear Power Plant, which shows the drastic parameters change during the system transient, a correlation having a reasonable degree of accuracy over a wide range is preferred, rather than that having accuracy for a specific range. It is required to have tangible insight about the effects of the CHF correlation to the LOCA analysis for the purpose of computer code development and nuclear regulation. The related research is further recommended. The purpose of this research is to obtain an insight and/or intuition about the above effect and to evaluate the selected CHF correlations. To achieve these purposes LOCA is analysed for the UL-JIN 3 and 4 nuclear power plant, the Korea Standard Type Nuclear Power Plant and the Loss of Flow Test (LOFT) L2-5 experiment is simulated using the RELAP5/MOD3.1 computer code for each selected CHF correlation. The selected correlations are the AECL-UO Lookup Table, adapted in RELAP5 code; the K110 CHF correlation, developed by KAERI; and the original W-3 CHF correlation, developed by L.S. Tong. LOFT is also simulated using the AECL-UO Lookup Table having the CHF multiplication factors 0.5 and 1.5, and then compared with the result of the original Lookup Table and the experiment result. In the LOCA analysis, the CHF correlations affect the magnitude of peak cladding temperatures, but does not seriously affect the occurrence points of time. The effect of each CHF correlation to the fuel cladding temperature behavior becomes apparent at the end of

  9. Study for Relation of Pressure and Aging Degradation during LOCA Test

    International Nuclear Information System (INIS)

    Kim, Jong Seog

    2013-01-01

    As result of this test, it was found that low pressure effect in aging was not significant compared with that of temperature. If temperature profile in LOCA test can satisfy the plant LOCA profile, no further analysis of pressure profile for aging degradation is necessary. For environmental qualification of electric equipment in containment building of nuclear power plant, LOCA test should be applied. During the LOCA test, temperature and pressure of LOCA chamber shall be controlled to meet a requirement of plant specific LOCA profile. It is general to keep LOCA test temperature and pressure above the plant specific LOCA profile. If the test temperature is lower than required profile in some time zone while it is higher in other time zone, calculation of total cumulated test temperature is required to compare with that of plant profile. Arrhenius equation can be applied for calculation of total temperature accumulation. If there is a deviation of pressure between test profile and plant specific profile, can we still use the same rule of temperature? Since the Arrhenius equation can't be applied to pressure, analysis of pressure effect to aging degradation is not easy. Study for relation of pressure and aging degradation during LOCA condition is described herein. To Study an aging degradation effect of pressure during LOCA test, comparison of IR during high LOCA pressure and low LOCA pressure were implemented. We expected low IR in high pressure because it contained a high concentration of oxygen which induces high aging degradation. Contrary to our expectation, IR of low pressure was lower than that of high pressure. It is assumed that high vibration of temperature profile to maintain the low pressure at high temperature induced supply of high enthalpy steam into LOCA chamber

  10. Sensitivity Study on Analysis of Reactor Containment Response to LOCA

    International Nuclear Information System (INIS)

    Chung, Ku Young; Sung, Key Yong

    2010-01-01

    As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

  11. Sensitivity Study on Analysis of Reactor Containment Response to LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Ku Young; Sung, Key Yong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2010-10-15

    As a reactor containment vessel is the final barrier to the release of radioactive material during design basis accidents (DBAs), its structural integrity must be maintained by withstanding the high pressure conditions resulting from DBAs. To verify the structural integrity of the containment, response analyses are performed to get the pressure transient inside the containment after DBAs, including loss of coolant accidents (LOCAs). The purpose of this study is to give regulative insights into the importance of input variables in the analysis of containment responses to a large break LOCA (LBLOCA). For the sensitivity study, a LBLOCA in Kori 3 and 4 nuclear power plant (NPP) is analyzed by CONTEMPT-LT computer code

  12. Development of Advanced Non-LOCA Analysis Methodology for Licensing

    International Nuclear Information System (INIS)

    Jang, Chansu; Um, Kilsup; Choi, Jaedon

    2008-01-01

    KNF is developing a new design methodology on the Non-LOCA analysis for the licensing purpose. The code chosen is the best-estimate transient analysis code RETRAN and the OPR1000 is aimed as a target plant. For this purpose, KNF prepared a simple nodal scheme appropriate to the licensing analyses and developed the designer-friendly analysis tool ASSIST (Automatic Steady-State Initialization and Safety analysis Tool). To check the validity of the newly developed methodology, the single CEA withdrawal and the locked rotor accidents are analyzed by using a new methodology and are compared with current design results. Comparison results show a good agreement and it is concluded that the new design methodology can be applied to the licensing calculations for OPR1000 Non-LOCA

  13. A synopsis of experimental activities on small-break LOCA

    International Nuclear Information System (INIS)

    Hein, D.

    1984-01-01

    Through reactor safety studies like WASH 1400 or the ''Deutsche Risiko-Studie'' the attention has turned from large break loss of coolant accidents to small breaks because of the high contribution of this type of accidents to core meltdown. But only after the TMI-2 accident were also the main activities in the experimental fields shifted world-wide to the small break LOCAs. Since TMI numerous research programs have either been finished or are underway. This review paper presents: a classification of the various types of transients according to break size; a discussion of major physical phenomena associated with a small break LOCA, and a description of a few selected research programs and the most important results achieved. (author)

  14. Analysis for Passive Safety Injection of IPSS in Various LOCAs

    International Nuclear Information System (INIS)

    Kim, Sangho; Chang, Soonheung

    2013-01-01

    The Fukushima accident shows US the possibility of accidents that are beyond a designed imagination. Lots of lessons can be shortly summarized into three issues. First of all, the original cause was the occurrence of a Station Black-Out (SBO). Even if engineers considered the possibility of a loss of offsite power enough to be managed, the failure of EDGs seemed to be unnoticed. The second is poor operation and accident management. They could not understand the overall system and did not check the availability of alternating systems. The third is the large release of radioactive materials outside the containment. Even if SBO occurred and the accident was not managed well, all the means must have prevented the large release out of containment. After that, lots of problems were pointed and numerous actions were carried out in each country. The representative proposals are AAC, additional physical barrier, bunker concept and large big tank. Integrated passive safety system (IPSS) was proposed as one of the solutions for enhancing the safety. IPSS can cope with a SBO and accidents with a SBO. IPSS has five functions which are passive decay heat removal, passive safety injection, passive containment cooling, passive in-vessel retention and filtered venting system. The results showed a high performance of removing decay heat through steam generator cooling by forming natural circulation in the primary circuit. The design concept of passive safety injection system (PSIS) consists of the injection line from integrated passive safety tank (IPST) to reactor vessel. The previous works were only focused on a double ended guillotine break LOCA in SBO. The purpose of this paper is to analyze the performance of PSIS in IPSS for various LOCAs by using MARS (Multi-dimensional Analysis of Reactor Safety) code. The simulated accidents were LOCAs which were accompanied with a SBO. The conditions of the LOCAs were varied only for the size of break. It shall show the capability of PSIS

  15. PWR cold-leg small break loca with faulty HPI

    International Nuclear Information System (INIS)

    Kumamaru, H.; Kukita, Y.

    1991-01-01

    The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

  16. Bulging of pressure tubes at hot spots under LOCA conditions

    International Nuclear Information System (INIS)

    Manu, C.; Shewfelt, R.S.W.; Wright, A.C.D.; Aboud, R.; Lau, J.H.K.; Sanderson, D.B.

    1996-01-01

    During certain postulated loss-of-coolant accidents (LOCA) in a CANDU reactor, some fuel channels can become highly voided within a very short time. Although the pressure tubes are heated mainly by convection and thermal radiation during the LOCA transient, additional heat flow occurs through the bearing pads that are in contact with the pressure tribe. This contact can lead to local hot spots and associated thermal stresses in the pressure tube wall. The two factors that affects the behavior of the pressure tubes during LOCA conditions are the internal pressure and the local heating. Although the effect of internal pressure and of axially uniform temperature has been studied elsewhere, the effect of the local heating on the pressure tube behavior has not been modelled before. This paper shows that the bulging of a pressure tube at a hot spot is the result of the thermal stresses that are developed in a pressure tube during a LOCA transient. To isolate the local heating effect from the internal pressure, a series of single-effect experiments was performed. In these experiments, sections of a CANDU pressure tube were subjected to local heating only. The thermal profile and the local deformation were measured function of time. To quantify the effect of the thermal stresses on the bulging of pressure tubes at hot spots and to develop numerical tools that can predict such bulging, finite element analyses were performed rising the ABAQUS finite element computer code. Use of the measured thermal profiles in the ABAQUS finite element analysis, resulted in very good agreement between the predicted and measured displacements. (author)

  17. Effect of oxygen in the simulated LOCA environments of the degradation of cable insulating materials

    International Nuclear Information System (INIS)

    Kusuma, Y.; Okada, S.; Itoh, M.; Yagi, T.; Yoshikawa, M.; Yoshida, K.; Machi, S.; Tamura, N.; Kawakami, W.

    1990-01-01

    Five kinds of insulating and jacketing materials for the cables used in nuclear power plants were exposed to various LOCA environments of both simultaneous and sequential methods using SEAMATE-II. Experimental conditions of the simultaneous LOCA tests were done at different radiation dose rate, steam temperature and amount of air added to the LOCA environments. The sequential tests consist of two stages, that is, pre-irradiation and subsequent steam/spray exposure. Pre-irradiation conditions and subsequent steam/spray exposure conditions of the sequential LOCA tests are systematically changed in order to find appropriate conditions which can bring about the degradation of same degree to those obtained for various simultaneous LOCA simulations. Tensile properties, insulating resistance and water sorption of the insulating materials exposed to various LOCA environments are measured and discussed. (author). 11 refs, 19 figs, 3 tabs

  18. An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

    Directory of Open Access Journals (Sweden)

    Cesare Frepoli

    2008-01-01

    Full Text Available Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

  19. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

    A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

  20. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  1. KUPOL-M code for simulation of the VVER's accident localization system under LOCA conditions

    International Nuclear Information System (INIS)

    Efanov, A.D.; Lukyanov, A.A.; Shangin, N.N.; Zajtsev, A.A.; Solov'ev, S.L.

    2004-01-01

    Computer code KUPOL-M is developed for analysis of thermodynamic parameters of medium within full pressure containment for NPPs with VVER under LOCA conditions. The analysis takes into account the effects of non-stationary heat-mass transfer of gas-drop mixture in the containment compartments with natural convection, volume and surface steam condensation in the presence of noncondensables, heat-mass exchange of the compartment atmosphere with water in the sumps. The operation of the main safety systems like a spray system, hydrogen catalytic recombiners, emergency core cooling pumps, valves and a fan system is simulated in KUPOL-M code. The main results of the code verification including the ones of the participation in ISP-47 International Standard Problem on containment thermal-hydraulics are presented. (author)

  2. Thematic network for a Phebus FPT1 international standard problem (THENPHEBISP)

    International Nuclear Information System (INIS)

    Clement, B.; Haste, T.; Krausmann, E.; Dickinson, S.; Gyenes, G.; Duspiva, J.; Rosa, F. de; Paci, S.; Martin-Fuertes, F.; Scholytssek, W.; Allelein, H.-J.; Guentay, S.; Arien, B.; Marguet, S.; Leskovar, M.; Sartmadjiev, A.

    2005-01-01

    The THENPHEBISP 2-year thematic network started in December 2001, and was concerned with OECD/CSNI International Standard Problem 46, itself based on the Phebus FPT1 core degradation/source term experiment. The aim was to assess the capability of computer codes to model in an integrated way the physical processes taking place during a severe accident in a pressurised water reactor, from the initial stages of core degradation, the fission product transport through the primary circuit and the behaviour of the released fission products in the containment. ISP-46, coordinated by IRSN/DRS Cadarache, attracted 33 participating organisations, from 23 countries and international bodies, who submitted 47 base case calculations and 21 best-estimate calculations, using 15 different codes. The thermal behaviour of the fuel bundle and the hydrogen production were generally well captured, and good agreement for the core final state could be obtained with a suitable choice of bulk fuel relocation temperature, however this is unlikely to be representative of all plant studies so sensitivity calculations are needed with the modelling in its current state. Total volatile fission product release was simulated, but its kinetics, and the overall modelling of semi-volatile, low-volatile and structural material release (Ag/In/Cd, Sn) needs improvement. Overall retention in the circuit is well predicted, but calculations underestimate deposits in the upper plenum and overestimate those in the steam generator, also the volatility of some elements could be better predicted. Containment thermal hydraulics and depletion rate of aerosols are well calculated, but with difficulties related to partition amongst the deposition mechanisms. Calculation of iodine chemistry in the containment turned out to be more difficult. Its quality strongly depends of the calculation of release and transport in the integral codes. The major difficulties are related to the existence of gaseous iodine in the

  3. Benchmark Calculations on Halden IFA-650 LOCA Test Results

    International Nuclear Information System (INIS)

    Ek, Mirkka; Kekkonen, Laura; Kelppe, Seppo; Stengaard, J.O.; Josek, Radomir; Wiesenack, Wolfgang; Aounallah, Yacine; Wallin, Hannu; Grandjean, Claude; Herb, Joachim; Lerchl, Georg; Trambauer, Klaus; Sonnenburg, Heinz-Guenther; Nakajima, Tetsuo; Spykman, Gerold; Struzik, Christine

    2010-01-01

    The assessment of the consequences of a loss-of-coolant accident (LOCA) is to a large extent based on calculations carried out with codes especially developed for addressing the phenomena occurring during the transient. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have not only generated a need to re-examine the LOCA safety criteria and to verify their continued validity, but also to confirm that codes show an appropriate performance especially with respect to high burnup phenomena influencing LOCA fuel behaviour. As part of international efforts, the OECD Halden Reactor Project program implemented a test series to address particular LOCA issues. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The Halden LOCA series, using high burnup fuel segments, contains test cases well suited for checking the ability of LOCA analysis codes to predict or reproduce the measurements and to provide clues as to where the codes need to be improved. The NEA Working Group on Fuel Safety, WGFS, therefore decided to conduct a code benchmark based on the Halden LOCA test series. Emphasis was on the codes' ability to predict or reproduce the thermal and mechanical response of fuel and cladding. Before starting the benchmark, participants were given the opportunity to tune their codes to the experimental system applied in the Halden LOCA tests. To this end, the data from the two commissioning runs were made available. The first of these runs went

  4. Review of RIA and LOCA criteria for WWER fuel

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    The RIA and LOCA fuel safety criteria are under revision in the international community of fuel suppliers, authorities and research organizations. The main criteria will be reviewed in the paper for WWER fuel. Experimental data on the fuel failure behaviour under reactivity-initiated accident (RIA) conditions produced in the last decade in French and Japanese test reactors indicated low failure enthalpy for high burnup fuel compared to fresh fuel. However the high burnup was not the only phenomenon influencing the fuel failure. The oxide scale on the external surface of the fuel rod, hydrogen content of the Zr cladding and the local hydriding seemed also be responsible for the failure at low enthalpy. Furthermore differences have been found between Western design fuel and Russian type WWER fuel. The burnup dependence of fuel failure for WWER fuel was found much less, probably due to the low oxidation during normal operational conditions compared to other PWRs. The recently published Vitanza and KAERI correlations for RIA failure enthalpy have been applied to 23 WWER tests. Experimental data from Russian IGR and BIGR reactors have been used. The calculations have shown that both burnup and cladding oxidation effects must be considered, however the pulse width dependence of failure enthalpy has not been confirmed. During loss of coolant accidents (LOCA) the peak cladding temperature and local oxidation criteria have to be met. The oxidation criterion is under discussion today in many laboratories. The AEKI carried out several experimental series with Zr1%Nb cladding used in WWER reactors. The paper will describe the main results of the tests and present the limit for ductile-brittle transition derived from ring compression test. The behaviour of Zr1%Nb (E110) and Zircaloy-4 claddings under LOCA conditions will be compared as well. (author)

  5. LOCA analysis of SCWR-M with passive safety system

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

    2013-06-15

    Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

  6. Bio-mechanical assessment toward throwing and lifting process of i-LOCA (Innovative Lobster Catcher)

    Science.gov (United States)

    Sudiarno, A.; Dewi, D. S.; Putri, M. A.

    2018-04-01

    Indonesia is the country rich in marine resource, one of which is lobster. East java, one of Indonesian province, especially in Region of Gresik and Lamogan, has very huge potential of lobster. Current condition shown that lobster catch by the fisherman mostly depend on lucky factor, which the lobster unintentionally trapped in fisherman’s fish net. By using this mechanism, the number of lobster catch cannot be optimum. Previous researches have produced two versions of i-LOCA, Innovative Lobster Catcher, a special tool for catching the lobster. Although produce more lobster catch, second version of i-LOCA still needs to be scrutinized, one of that is bio-mechanical assessment. The second version of i-LOCA still has no tool to ease throwing and lifting it into the sea. This condition cause Musculoskeletal Disorder (MSD) toward the fisherman. This research perform bio-mechanical assessment toward throwing and lifting process in order to suggest improvement for i-LOCA as the third version. Based on body moment calculation, we found that throwing and lifting process of third version of i-LOCA, each was 3 times and 2 times better than second version of i-LOCA. Meanwhile, Rapid Entire Body Assessment (REBA) score of throwing and lifting process for third version of i-LOCA can be reduced by 5 points compared to second version of i-LOCA.

  7. Comparison of models discribing cladding deformations during LOCA

    International Nuclear Information System (INIS)

    Chakraborty, A.K.; Zipper, R.

    1981-05-01

    This report compares the important models for the determination of cladding deformations during LOCA. In addition to the comparisons of underlying assumptions of different models the same is done for the coefficients applied for the models. In order to assess the predictive capability of the models the calculated results are compared with the experimental results of the individual claddings. It was found out that the results of temperature ramp tests could be calculated better than that of the pressure ramp tests. The calculations revealed that even with the simplified assumption of the model used in TESPA the agreement of the calculated results with those of model NORA was relatively good. (orig.) [de

  8. LOCA scenario tests of irradiated fuel rod specimens

    International Nuclear Information System (INIS)

    Scott, Harold

    2004-01-01

    Full text: The NRC's cladding performance program at Argonne National Laboratory (ANL) is testing fueled high-burnup segments subjected to LOCA integral phenomena. The data are provided to NRC and the nuclear industry for their independent assessment of the adequacy of licensing criteria for LOCA events. The tests are being conducted with high-burnup 30 cm segments from Limerick (9x9 Zry-2) and H.B. Robinson (15x15 Zry-4) reactors. Prior to testing, sibling samples are characterized with respect to fuel morphology, fuel-cladding bond, cladding oxide layer thickness, hydrogen content and high-temperature steam oxidation kinetics. Specimens that survive quench are subjected to four-point bend tests, followed by local diametral compression tests. The retention of post-quench ductility is a more limiting requirement than surviving thermal stresses during quench. Companion tests are conducted with unirradiated cladding to generate baseline data for comparison with the high-burnup fuel results. LOCA integral tests have the following sequential steps: stabilization of temperature, internal pressure and steam flow at 300 C, ramping of temperature (∼5C/s) through ballooning and burst to 1204 C, hold at 1204 C for 1-5 minutes, slow-cooling (∼3C/s) to 800 C, and water quenching at ∼800C. Two high-burnup tests were completed in 2002 with Limerick BWR rod segments: ramp to burst in argon followed by slow cooling; and the LOCA test with 5-minute hold time at 1204 C, followed by slow cooling. With the exception of burst-opening shape, results for burst temperature, burst pressure, burst length, and ballooning strain profile are more similar to, than different from, results for unirradiated Zry-2 cladding exposed to the same time-temperature history. The 3rd Limerick test with quench was performed in December 2003, and a 4th Limerick test was performed in March 2004. Tests on high-burnup Robinson PWR fuel segments are scheduled to begin in June 2004. The presentation points

  9. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R.; Erixon, S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B. [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B. [RSA Technologies, Visat, CA (United States)

    1996-09-01

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs.

  10. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    International Nuclear Information System (INIS)

    Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.

    1996-09-01

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs

  11. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  12. Consistent Posttest Calculations for LOCA Scenarios in LOBI Integral Facility

    Directory of Open Access Journals (Sweden)

    F. Reventós

    2012-01-01

    Full Text Available Integral test facilities (ITFs are one of the main tools for the validation of best estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment-scaled conditions in a full NPP. The LOBI was a single plus a triple-loop (simulated by one loop test facility electrically heated to simulate a 1300 MWe PWR. The scaling factor was 712 for the core power, volume, and mass flow. Primary and secondary sides contained all main active elements. Tests were performed for the characterization of phenomenologies relevant to large and small break LOCAs and special transients in PWRs. The paper presents the results of three posttest calculations of LOBI experiments. The selected experiments are BL-30, BL-44, and A1-84. They are LOCA scenarios of different break sizes and with different availability of safety injection components. The goal of the analysis is to improve the knowledge of the phenomena occurred in the facility in order to use it in further studies related to qualifying nodalizations of actual plants or to establish accuracy data bases for uncertainty methodologies. An example of procedure of implementing changes in a common nodalization valid for simulating tests occurred in a specific ITF is presented along with its confirmation based on posttests results.

  13. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  14. SB LOCA analyses for Krsko Full Scope Simulator verification

    International Nuclear Information System (INIS)

    Prosek, A.; Parzer, I.; Mavko, B.

    2000-01-01

    Nuclear power plant simulators are intended to be used for training and maintaining competence to ensure safe, reliable operation of nuclear power plants throughout the world. The simulator shall be specified to a reference unit and its performance validation testing shall be provided. In this study a small-break loss-of-coolant accident (SB LOCA) response of Krsko nuclear power plant (NPP) was calculated for full scope simulator verification. The investigation included five cases with varying the break size in the cold leg of reactor coolant system. The plant specific and verified RELAP5/MOD2 model of Krsko nuclear power plant (NPP), developed in the past for 1882 MWt power, was adapted for 2000 MWt power (cycle 17) including the model for replacement steam generators. The results showed that the plant system response to breaks with small break area was slower compared to breaks with larger break area. The core heatup occurred in most of the cases analyzed. The acceptance criteria for emergency core cooling system were also met. The predicted results of the SB LOCA analysis for Krsko NPP suggest that they may be used for verification of the Krsko Full Scope Simulator performance. (author)

  15. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    International Nuclear Information System (INIS)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

    2017-01-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  16. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

    Mann, C.A.; Hindle, E.D.; Parsons, P.D.

    1982-04-01

    Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 1500 0 C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

  17. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

    The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

  18. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

    Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

    1991-10-01

    The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

  19. Upper-bound fission product release assessment for large break LOCA in CANFLEX bundle reactor core

    International Nuclear Information System (INIS)

    Oh, Duk Ju; Lee, Kang Moon

    1996-07-01

    Quarter-core gap inventory assessment for CANDU-6 reactor core loaded with CANFLEX fuel bundles has been performed as one of the licensing safety analyses required for 24 natural uranium CANFLEX bundle irradiation in CANDU-6 reactor. The quarter-core gap inventory for the CANFLEX bundle core is 5 - 10 times lower than that for the standard bundle core, depending on the half-life of the isotope. The lower gap inventory of the CANFLEX bundle core is attributed to the lower linear power of the CANFLEX bundle compared with the standard bundle. However, the whole core total inventories for both the CANFLEX and standard bundle cores are nearly the same. The 6 - 8 times lower upper-bound fission product releases of the CANFLEX bundle core for large break LOCA than those of the standard bundle core imply that the loading of 24 natural uranium CANFLEX bundles would improve the predicted consequences of the postulated accident described in the Wolsung 2 safety report. 2 tabs., 6 figs., 3 refs. (Author)

  20. Status of efforts to evaluate LOCA frequency estimates using combined PRA and PFM approaches

    International Nuclear Information System (INIS)

    Wilkowski, G.; Rudland, D.; Tregoning, R.; Scott, P.

    2002-01-01

    The risk-informed reevaluation of 10 CFR 50.46 (along with Appendix K and GDC 35), the emergency core cooling system (ECCS) requirements, utilizes loss of coolant accident (LOCA) initiating event frequencies to evaluate the technical basis for potential related rule changes. A longer-term effort is considering redefining the maximum design basis pipe break size for sizing the ECCS system. In the past few years, the U.S. Nuclear Regulatory Commission (NRC) has utilized NUREG/CR-5750 pipe-break LOCA estimated for initiating event frequencies. However, several failure mechanisms have recently emerged at plants which have not been evident within the service period covered by the NUREG/CR-5750 estimates. The concern is that these and other potential aging-related mechanisms may not be adequately represented within the NUREG/CR-5750 LOCA estimates. Additionally, LOCAs can occur from failure of active components (e.g. safety relief valves, reactor coolant pump seals, etc.) and other non-pipe break passive failures (e.g. steam generator tubes). The LOCA contributions from these additional sources must also be considered in deciding the design basis break size. The LOCA estimates must also attempt to capture expected future changes in the LOCA frequencies so that the estimates are pertinent up through the end of the license renewal period. (orig.)

  1. ALARM, Thermohydraulics of BWR with Jet Pumps During LOCA

    International Nuclear Information System (INIS)

    Araya, F.; Akimoto, M.

    1985-01-01

    1 - Nature of physical problem solved: ALARM-B2 which is an improved version of ALARM-B1 is a computer program to analyze thermo-hydraulic phenomena of BWR during a blowdown period under a large-break loss-of-coolant accident condition with special emphasis on the heat transfer phenomena in the core region. 2 - Method of solution: A so called volume-junction method is used to present fluid conservations. The primary system is divided into a number of special elements called 'control-volumes'. The system of partial differential equations describing fluid conservations for a stream-tube are integrated over a number of control volumes. The resulting set of simultaneous differential equations that is based on the assumptions of one-dimensional, homogeneous and thermal- equilibrium flow is linearized and solved for a small time increment by a simple explicit numerical technique. The one-dimensional heat conduction equations describing temperature profiles within solid material are written in finite difference forms which are linearized and solved by the Crank-Nicholson implicit method. In order to simulate the blowdown heat transfer phenomena, the code has correlation packages for heat transfer coefficient and critical heat flux. The heat generation in the core is given by a point reactor kinetics model with six groups of delayed neutrons and decay of eleven groups of fission products and actinides. The solution technique of the reactor kinetics is based on the Runge-Kutta method. ALARM-B2 has the models to simulate various components incorporated in BWRs such as jet pumps, recirculation pumps, steam separators, valves, and so on. The discharge and injection systems are modeled by leak and fill systems, respectively. 3 - Restrictions on the complexity of the problem: As this has been developed to simulate a blowdown thermo-hydraulic transient during a large break LOCA, users must pay attention when applying the code to any medium or small break LOCAs or to later phases

  2. ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-06-01

    Full Text Available ABSTRAK ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA. Kecelakaan yang diakibatkan oleh kehilangan pendingin (loss of coolant accident / LOCA dari sistem reaktor merupakan kejadian dasar desain yang tetap diantisipasi dalam desain reaktor daya yang mengadopsi teknologi Generasi II hingga IV. LOCA ukuran kecil (small break LOCA memiliki dampak yang lebih signifikan terhadap keselamatan dibandingkan LOCA ukuran besar (large break LOCA seperti terlihat pada kejadian Three-Mile Island (TMI. Fokus makalah adalah pada analisis small break LOCA pada reaktor daya maju Generasi III+ yaitu AP1000 dengan mensimulasikan tiga kejadian pemicu yaitu membukanya katup Automatic Depressurization System (ADS secara tak disengaja, putusnya salah satu pipa Direct Vessel Injection (DVI secara double-ended, dan putusnya pipa lengan dingin dengan diameter bocoran 10 inci. Metode yang digunakan adalah simulasi kejadian pada model AP1000 yang dikembangkan secara mandiri menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Dampak yang ingin dilihat adalah kondisi teras selama terjadinya small break LOCA yang terdiri dari pembentukan mixture level dan transien temperatur kelongsong bahan bakar. Hasil simulasi menunjukkan bahwa mixture level untuk semua kejadian small break LOCA berada di atas tinggi teras aktif yang menunjukkan tidak terjadinya core uncovery. Adanya mixture level berpengaruh pada transien temperatur kelongsong yang menurun dan menunjukkan pendinginan bahan bakar yang efektif. Hasil di atas juga identik dengan hasil perhitungan program lain yaitu NOTRUMP. Keefektifan pendinginan teras juga disebabkan oleh berfungsinya injeksi pendingin melalui fitur keselamatan pasif yang menjadi ciri reaktor daya AP1000. Secara keseluruhan, hasil analisis menunjukkan model AP1000 yang telah dikembangkan dengan RELAP5 dapat digunakan untuk keperluan analisis kecelakaan dasar desain pada reaktor daya maju AP1000. Kata kunci: analisis

  3. Best estimate small break LOCA analysis for KNGR SIS optimization

    International Nuclear Information System (INIS)

    Song, JIn Ho; Lim, Hong Sik; Bae, Kyoo Hwan; Lee, Joon

    1996-01-01

    The KNGR has an advanced ECCS design feature which employs four mechanically-separated SI trains where each train consisting of one HPSI pump and one SIT injects ECC water directly into the reactor vessel downcomer annulus. To demonstrate that the KNGR ECCS design features meet the EPRI ALWR requirements of no core uncovery for a break of up to 6 inch diameter, small break LOCA cases with various break sizes were analyzed using a best-estimate analytical procedure. Two kinds of break locations are considered: cold leg and DVI line breaks. It was observed that the KNGR ECCS design can tolerate a cold leg break of up to 10 inches with no core uncovery. However, since DVI line break with 6 inch diameter undergoes slight core uncovery, further investigation is required for KNGR SIS optimization

  4. A comparison of LOCA analysis using SMOKIN and CERBERUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Younis, M H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Gaboury, G [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    This paper presents the results of a comparison of the analyses of a postulated Loss of Coolant Accident (LOCA) in Pickering NGS A reactors using the two neutron kinetics codes SMOKIN and CERBERUS. Both codes have been used to simulate the space-time neutronic behaviour of CANDU-PHWR reactors. The main objective of the present study is to evaluate the accuracy with which SMOKIN can predict power transients compared to CERBERUS. The comparison shows that the two codes produce similar bulk power and reactivity transients. However, SMOKIN was found to overestimate the power transient (relative to CERBERUS) in some regions of the core, which is indicative of the spatial differences between the two codes. It was demonstrated that part of this overestimate is due to the use of reaction-rate averaged fuel properties in SMOKIN, compared to instantaneous fuel properties in CERBERUS. (author). 5 refs., 3 tabs., 6 figs.

  5. Measurement of mist cooling of PWR during LOCA by LDA

    International Nuclear Information System (INIS)

    Lee, S.L.; Sheen, H.J.; Issapour, I.

    1985-01-01

    The prediction of temperature distribution and heat transfer within rod bundles during the refill and reflood phase of a LOCA (loss of coolant accident) is of critical importance for determining the location and size of blockages due to clad deformation in a pressurized water reactor (PWR). Mist cooling by small droplets generated from large droplets on hitting grid spacers has been suggested as one of the most important heat transfer mechanisms which are responsible for the development of this temperature transient. The questions to be asked are whether such small droplets indeed exist and, if so, how are they related to the cooling of the fuel rods. Hereby reported is the result of a direct experimental investigation on these questions by a special laser-Doppler anemometry (LDA) particle sizing technique together with temperature measurements of the rod claddings and flow in the subchannel

  6. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

    Full Text Available This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER. The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model, represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

  7. BWR 200 % recirculation pump suction line break LOCA tests, RUNs 942 and 943 at ROSA-III without HPCS

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Koizumi, Yasuo

    1986-03-01

    This report presents the experimental results of RUNs 942 and 943 in ROSA-III program, which are 200 % recirculation pump suction line break LOCA tests with assumption of HPCS failure. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, ECCS's, and steam and feedwater systems. Effects of initial core void distribution and other fluid conditions on overall LOCA phenomena with special interest on transient core cooling phenomena were investigated by comparing the present test results with those of RUN 926, a 200 % suction line break test with standard initial fluid conditions. The initial core outlet quality was changed between 5 % and 43 %. As conclusions, (1) the initial lower core flow and higher void fraction affected significantly the core cooling conditions and resulted in earlier and higher PCT. (2) The lower plenum flashing temporarily contributed to cool down the core. (3) Flashing of remained hot water in the feedwater line affected slightly the pressure response and delayed the actuation of LPCI by 11 seconds. (4) The whole core was completely cooled down within 104 seconds after the LPCI actuation in these large break tests. (author)

  8. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Directory of Open Access Journals (Sweden)

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  9. Accomplishments of LOCA/ECCS experimental research at Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Murao, Yoshio; Koizumi, Yasuo

    1984-01-01

    Japan Atomic Energy Research Institute has investigated loss-of-coolant accident (LOCA)/emergency core cooling system (ECCS) from 1970. Major results of the LOCA/ECCS research are summarized in this report. ROSA-II program was LOCA/ECCS research for a pressurized water reactor (PWR) and ROSA-III program was for a boiling water reactor (BWR). The both test facilities were scaled at approximately 1/400 of the respective reference PWR and BWR. Large scale reflood test is research on reflood phenomena during a large break LOCA of PWR. The test facility is scaled at approximately 1/20 of the reference PWR and the research is still being continued. (author)

  10. Taipower's approach in development of in-house LOCA analysis capability

    International Nuclear Information System (INIS)

    Wang, L.C.

    2004-01-01

    One lesson learned from the Three Mile Island (TMI) accident was the analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or nuclear fuel suppliers for small break Loss Of Coolant Accident (LOCA) analysis for compliance with appendix K to 10CFR50 should be revised, so, a technology transfer program and a training program of a new LOCA analysis methodology for Taipower's engineers is briefly described in this paper. Also, an other lesson learned from the TMI accident was the plant-specific calculations using NRC-approved models for small-break LOCA to show compliance with 10CFR50.46 should be submitted for NRC approval, so, a study by Taiwan Power Company (TPC) under the guidance of Yankee Atomic Electric Company (YAEC) has been undertaken to perform this analysis for Maanshan nuclear power plant. The results of the 4 inch line break LOCA analysis is described in this paper. (author)

  11. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Muhammad Subekti

    2009-01-01

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  12. SCRELA, LOCA Analysis of Super-Critical Light-Water Reactors

    International Nuclear Information System (INIS)

    Lee, J.H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    Description of program or function: LOCA Analysis Code for the Supercritical-Water Cooled Reactor. - Blowdown Module: Calculation of the Blowdown Phase and Refill Phase. - Reflood Module: Calculation of the Reflood Phase

  13. LWR fuel cladding deformation in a LOCA and its interaction with the emergency core cooling

    International Nuclear Information System (INIS)

    Erbacher, F.J.

    1982-01-01

    The paper summarizes research results of out-of-pile burst tests, in-pile bursts tests, out-of-pile flooding tests and modeling work on fuel behavior in a LOCA performed at KfK: The dominant phenomena of the cladding deformation and failure have been clarified by experiments and can be modeled by computer codes. The burst and flooding tests performed up to now suggest that the coolability of the core under LOCA conditions can be maintained. (orig.) [de

  14. The blowdown, refill and reflood phase during a LOCA. Survey of the main physical phenomena

    International Nuclear Information System (INIS)

    Reocreux, M.

    1980-05-01

    In this paper, the main physical phenomena occuring during a LOCA are reviewed. They are presented in a chronological order. For each phenomena, a detailed physical description is given followed by the review of the general modelling problems. For some of these phenomena, modelling details are given for critical flow, for two-phase flow and heat transfer, for critical heat flux and post critical heat flux heat transfer, for reflood and rewet heat transfer and in the survey on LOCA computation codes

  15. Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

    International Nuclear Information System (INIS)

    Casal, V.; Malang, S.; Rust, K.

    1982-10-01

    For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

  16. Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

    International Nuclear Information System (INIS)

    Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick; Rudland, Dave

    2010-12-01

    The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

  17. Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

    2010-01-01

    To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

  18. Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

    International Nuclear Information System (INIS)

    Liang, Thomas K.S.; Chou, Ling-Yao; Zhang, Zhongwei; Hsueh, Hsiang-Yu; Lee, Min

    2011-01-01

    Highlights: → A new LOCA licensing methodology (DRHM, deterministic-realistic hybrid methodology) was developed. → DRHM involves conservative Appendix K physical models and statistical treatment of plant status uncertainties. → DRHM can generate 50-100 K PCT margin as compared to a traditional Appendix K methodology. - Abstract: It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.

  19. Development of the PRO-LOCA Probabilistic Fracture Mechanics Code, MERIT Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Scott, Paul; Kurth, Robert; Cox, Andrew; Olson, Rick (Battelle Columbus (United States)); Rudland, Dave (Nuclear Regulatory Commission (United States))

    2010-12-15

    The MERIT project has been an internationally financed program with the main purpose of developing probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code named PRO-LOCA. The principal objective of the project has been to develop probabilistic models for piping failure of nuclear components and to include these models in a probabilistic code. The MERIT program has produced a code named PRO-LOCA with the following features: - Crack initiation models for fatigue or stress corrosion cracking for previously unflawed material. - Subcritical crack growth models for fatigue and stress corrosion cracking for both initiated and pre-existing circumferential defects. - Models for flaw detection by inspections and leak detection. - Crack stability. The PRO-LOCA code can thus predict the leak or break frequency for the whole sequence of initiation, subcritical crack growth until wall penetration and leakage, instability of the through-wall crack (pipe rupture). The outcome of the PRO-LOCA code are a sequence of failure frequencies which represents the probability of surface crack developing, a through-wall crack developing and six different sizes of crack opening areas corresponding to different leak flow rates or LOCA categories. Note that the level of quality assurance of the PRO-LOCA code is such that the code in its current state of development is considered to be more of a research code than a regulatory tool.

  20. Safety Significance of the Halden IFA-650 LOCA Test Results

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Petit, Marc; Hozer, Zoltan; Kelppe, Seppo; Khvostov, Grigori; Hafidi, Biya; Therache, Benjamin; Heins, Lothar; Valach, Mojmir; Voglewede, John; Wiesenack, Wolfgang

    2010-01-01

    The safety criteria for loss-of-coolant accidents were defined to ensure that the core would remain coolable. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have generated a need to re-examine these criteria and to verify their continued validity. As part of international efforts to this end, the OECD Halden Reactor Project program implemented a LOCA test series. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, IRSN, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The fourth test of the series, IFA-650.4 conducted in April 2006, caused particular attention in the international nuclear community. The fuel used in the experiment had a high burnup, 92 MWd/kgU, and a low pre-test hydrogen content of about 50 ppm. The test aimed at and achieved a peak cladding temperature of 850 deg. C. The rod burst occurred at 790 deg. C. The burst caused a marked temperature increase at the lower end and a decrease at the upper end of the system, indicating that fuel relocation had occurred. Subsequent gamma scanning showed that approximately 19 cm of the fuel stack were missing from the upper part of the rod and that fuel had fallen to the bottom of the capsule. PIE at the IFE-Kjeller hot cells corroborated this evidence of substantial fuel relocation. The fact that fuel dispersal could occur upon ballooning and burst, i.e. at cladding temperatures as low as 800 deg. C and thus far lower than the temperature entailed by the current 1200 deg. C / 17% ECR limit, caused concern. The

  1. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  2. An analysis of LOCA sequences in the development of severe accident analysis DB

    International Nuclear Information System (INIS)

    Choi, Young; Park, Soo Yong; Ahn, Kwang-Il; Kim, D.H.

    2006-01-01

    Although a Level 2 PSA was performed for the Korean Standard Power Plants (KSNPs), and it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results. At present, KAERI is calculating the severe accident sequences intensively for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviours. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by knowledge-base technique, and the expected plant behaviour. The plant model used in this paper is oriented to the case of LOCAs related severe accident phenomena and thus can simulate the plant behaviours for a severe accident. Therefore the developed system may play a central role as an information source for decision-making for a severe accident management, and will be used as a training simulator for a severe accident management. (author)

  3. Sampling based uncertainty analysis of 10% hot leg break LOCA in large scale test facility

    International Nuclear Information System (INIS)

    Sengupta, Samiran; Kraina, V.; Dubey, S. K.; Rao, R. S.; Gupta, S. K.

    2010-01-01

    Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between 5 th and 95 th percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure

  4. An assessment of post-LOCA radiolytic generation of hydrogen in reactor containment of Indian PHWRs

    International Nuclear Information System (INIS)

    Bose, H.; Shah, G.C.; Dutta, S.

    2002-01-01

    Full text: An event-wise assessment has been carried out for the 220 MWe Indian PHWRs of standardized design, to estimate the post-LOCA release of radiolytic hydrogen inside reactor containment, in absence of steam-zirconium reaction. The assessment is based on (i) the dissolved hydrogen concentration build-up in water corresponding to the decaying gamma dose profile and (ii) the rate of concentration dependent mass-transfer of hydrogen from water to gas-space. It is observed that the total radiolytic hydrogen released is about three times less than that obtained by the conventional method of calculation which assumes the radiolytic yield of hydrogen to be equal to the primary yield G(H 2 ) = 0.44 molecules per 100 eV. It is also seen that a major part (∼90 %) of the total release is due to the spillage of fission product irradiated suppression pool water flowing through the core, followed by moderator and suppression pool surface releases respectively

  5. Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data

    International Nuclear Information System (INIS)

    2012-08-01

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and cooperative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. Two RD-14M small break loss of coolant accident (SBLOCA) tests, simulating HWR LOCA behaviour, conducted by Atomic Energy of Canada Ltd (AECL), were selected for this validation project. This report provides a comparison of the results obtained from eight participating organizations from six countries (Argentina, Canada, China, India, Republic of Korea, and Romania), utilizing four different computer codes (ATMIKA, CATHENA, MARS-KS, and RELAP5). General conclusions are reached and recommendations made.

  6. Calculation of the frequency of excedence in Full Spectrum LOCA by FSR; Calculo de la Frecuencia de excedencia en Full Spectrum LOCA mediante metodologia ISA

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Magan, J. J.; Queral Salazar, C.; Sanchez Perea, M.

    2012-07-01

    In this application LOCA sequences was taken into account the uncertainty in the size of rupture and the operator action times as cooling and depressurization through steam generators. The simulations were performed using the tool SCAIS, dynamically coupled with MAAP code.

  7. Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

    We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

  8. VVER-1000 small-medium break LOCAs predictions by ASTEC

    International Nuclear Information System (INIS)

    Georgieva, J.; Stefanova, A.; Atanasova, B.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

    2005-01-01

    This paper deals with an assessment of ASTEC1.1v0 code in the simulation of small and medium break LOCAs (ranging from 30mm up to 70mm equivalent diameter). The reference power plant for this analysis is a VVER-1000/V320 (e.g. Units 5 and 6 at Kozloduy NPP). A preliminary comparison with MELCOR and RELAP-SCDAP severe accident codes will be discussed. This investigation has been performed in the framework of the SARNET project (under the Euratom 6 th framework program) by the FoBAUs group (Forum of Bulgarian ASTEC users). The FoBAUs group aims at the validation of the ASTEC code in the field of severe accidents. Future activities will target the ASTEC capability (as a PSA-level 2 tool) to simulate a large range of reactor accident scenarios with intervention of safety systems (either passive systems or operated by operators). The final target is to assess Severe Accident Management (SAM) procedures for VVER-1000 reactors. The ASTEC1.1v0 code version here used is the one released in June 2004 by the French IRSN (Institut de Radioprotection et de Surete Nucleaire) and the German GRS (Gesellschaft ReactorSicherheit mbH). (author)

  9. HCCR TBS LOCA and ICE into small confined volume

    International Nuclear Information System (INIS)

    Jin, Hyung Gon; Ahn, Mu-Young

    2016-01-01

    KAERI has participated in the development of HCCR (Helium Cooled Ceramic Reflector) TBS (Test Blanket System) as a member of the KO TBM Team. Conceptual design review of this system had been performed in 2015 and after resolving the chits, the final approval was achieved in March 2016. This safety issue is one of the category II chits in the CDR and resolution strategy was already approved, however, safety analysis should be done until PDR (Preliminary Design Review). In this paper, model and nodalization for the accident are given and preliminary result is included. Nominal design pressure of HCS loop is 8 MPa, therefore, as indicated in the figure below. During the break of cooling pipe between TBM and Shield, the high pressure coolant will ingress to the 'interspace' between TBM, Shield and Frame. The coolant will be released through the front gaps between TBM and Frame towards VV primary vacuum. Accident analysis about HCCR TBS LOCA and ICE into small confined volume has been done successfully. Inverspace volume is compatibly small volume for 8MPa helium loop rupture, which causes fast pressure build-up the space but it decrease within 10 seconds. It is expected that other type of TBM has almost the same behavior

  10. Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

    Directory of Open Access Journals (Sweden)

    Lázaro Lima

    2011-12-01

    Full Text Available “Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

  11. Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions

    International Nuclear Information System (INIS)

    Pettersson, Kjell; Chung, Haijung; ); Billone, Michael; Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Hache, George; Papin, Joelle; Heins, Lothar; Hozer, Zoltan; In de Betou, Jan; Kelppe, Seppo; Mayer, Ralph; Scott, Harold; Voglewede, John; Sonnenburg, Heinz; Sunder, Sham; Valach, Mojmir; Vrtilkova, Vera; Waeckel, Nicolas; Wiesenack, Wolfgang; Zimmermann, Martin

    2009-01-01

    The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the current understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burn-up and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including as regards experimental approaches and interpretation and the use of experimental data relevant for safety. In 1986, a working group of the NEA Committee on the Safety of Nuclear Installations (CSNI) issued a state-of-the-art report on water reactor fuel behaviour in design-basis accident (DBA) conditions. The 1986 report was limited to the oxidation, embrittlement and deformation of pressurised water reactor (PWR) fuel in a loss-of-coolant accident (LOCA). Since then, considerable experimental and analytical work has been performed, which has led to a broader and deeper understanding of LOCA-related phenomena. Further, new cladding alloys have been produced, which might behave differently than the previously used Zircaloy-4, both under normal operating conditions and during transients. Compared with 20 years ago, fuel burn-up has been significantly increased, which requires extending the LOCA database in order to cover the high burnup range. There was also a clear need to address LOCA performance for reactor types other than PWRs. The present report has been prepared by the WGFS and covers the following technical aspects: - Description of different LOCA scenarios for major types of reactors: BWRs, PWRs, VVERs and to a lesser extent CANDUs. - LOCA phenomena: ballooning, burst, oxidation, fuel relocation and possible fracture at quench. - Details of high-temperature oxidation behaviour of various cladding materials. - Metallurgical phase change, effect of hydrogen and oxygen on residual cladding ductility. - Methods for LOCA testing, for example two-sided oxidation and ring compression for ductility, and integral quench test for

  12. Utilizing elements of the CSAU phenomena identification and ranking table (PIRT) to qualify a PWR non-LOCA transients system code

    Energy Technology Data Exchange (ETDEWEB)

    Greene, K.R.; Fletcher, C.D.; Gottula, R.C.; Lindquist, T.R.; Stitt, B.D. [Framatome ANP, Richland, WA (United States)

    2001-07-01

    Licensing analyses of Nuclear Regulatory Commission (NRC) Standard Review Plan (SRP) Chapter 15 non-LOCA transients are an important part of establishing operational safety limits and design limits for nuclear power plants. The applied codes and methods are generally qualified using traditional methods of benchmarking and assessment, sample problems, and demonstration of conservatism. Rigorous formal methods for developing code and methodology have been created and applied to qualify realistic methods for Large Break Loss-of-Coolant Accidents (LBLOCA's). This methodology, Code Scaling, Applicability, and Uncertainty (CSAU), is a very demanding, resource intensive, process to apply. It would be challenging to apply a comprehensive and complete CSAU level of analysis, individually, to each of the more than 30 non-LOCA transients that comprise Chapter 15 events. However, certain elements of the process can be easily adapted to improve quality of the codes and methods used to analyze non- LOCA transients. One of these elements is the Phenomena Identification and Ranking Table (PIRT). This paper presents the results of an informally constructed PIRT that applies to non-LOCA transients for Pressurized Water Reactors (PWR's) of the Westinghouse and Combustion Engineering design. A group of experts in thermal-hydraulics and safety analysis identified and ranked the phenomena. To begin the process, the PIRT was initially performed individually by each expert. Then through group interaction and discussion, a consensus was reached on both the significant phenomena and the appropriate ranking. The paper also discusses using the PIRT as an aid to qualify a 'conservative' system code and methodology. Once agreement was obtained on the phenomena and ranking, the table was divided into six functional groups, by nature of the transients, along the same lines as Chapter 15. Then, assessment and disposition of the significant phenomena was performed. The PIRT and

  13. Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong Kwan, E-mail: jksuh@khnp.co.kr [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kim, Jae Won; Kwon, Sun Guk; Lee, Jae Yong [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Cho, Hyoung Kyu; Park, Goon Cherl [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

    2015-08-15

    Highlights: • In-vessel downstream effect tests were performed in the presence of LOCA-generated debris. • Available driving heads under each LOCA scenario were verified using experimental data. • Fibrous debris was prepared to satisfy the length distribution obtained from the bypass test. • Limiting test conditions were identified through sensitivity studies. - Abstract: Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

  14. LOFT/L3-, Loss of Fluid Test, 7. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the seventh in the NRC L3 Series of small-break LOCA experiments. A 2.5-cm (10-in.) cold-leg non-communicative-break LOCA was simulated. The experiment was conducted on 20 June 1980

  15. Fuel Behaviour at High During RIA and LOCA Accidents

    International Nuclear Information System (INIS)

    Barrio del Juanes, M. T.; Garcia Cuesta, J. C.; Vallejo Diaz, I.; Herranz Puebla

    2001-01-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs

  16. Large LOCA accident analysis for AP1000 under earthquake

    International Nuclear Information System (INIS)

    Yu, Yu; Lv, Xuefeng; Niu, Fenglei

    2015-01-01

    Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

  17. THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2001-01-01

    1 - Description of problem or function: THYDE-P1 analyzes the behaviour of LWR plants in response to various disturbances, including the thermal hydraulic transient following a break of the primary coolant pipe system, generally referred to as a loss-of-coolant-accident (LOCA). LOCA can be considered as the most critical condition for testing the methods and models for plant dynamics, since thermal hydraulic conditions in the system change drastically during the transient. THYDE-P is capable of a complete LOCA calculation from start to complete reflooding of the core by subcooled water. The program performs steady-state adjustment, which is complete in the sense that the steady state obtained is a set of exact solutions of all the transient equations without time derivatives, not only for plant hydraulics but also for all the other phenomena in the simulation of a PWR plant. THYDE-P2 contains among others the following improvements over THYDE-P1: (1) not only the mass and momentum equations but also the energy equation are included in the non-linear implicit scheme; (2) the valve model is implemented; (3) the relaxation equation for void fraction is theoretically derived; (4) vectorized programming is implemented; (5) both EM (evaluation mode) and BE (best estimate) calculations are possible. THYDE-W is an improved version of THYDE-P2 and contains the following additional features: (a) analysis of multiple number of disjoint loops is possible; (b) a control system simulation model is included; (c) the trip model has been improved; (d) heavy water is allowed as coolant; (e) the effect of drift flux is accounted for in the steady state calculation; (f) boron transport is included; (g) to obtain steady state loop heat balance, the option of adjusting the enthalpy distribution is prepared included in addition to that of adjusting heat exchanger areas; (h) to obtain steady state pressure distribution, three other options are prepared in addition to the original ones

  18. Cobalt-60 simulation of LOCA [loss of coolant accident] radiation effects

    International Nuclear Information System (INIS)

    Buckalew, W.H.

    1989-07-01

    The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs

  19. About criteria of inadmissible embrittlement of zirconium fuel cladding during LOCA in the PWRs

    International Nuclear Information System (INIS)

    Osmachkin, V.S.

    1999-01-01

    According the licensing procedures the designers of the PWRs reactor have to prove the meeting of special safety requirements. One criteria on effectiveness of the Emergency Core Cooling System is not to exceeding some limited conditions of the fuel cladding during LOCA accidents (typical example T m ax o C, ECR<0,17 and oth.). The damage of fuel element in the core during LOCA is caused by the oxidation of the cladding, its embrittlement and thermal shock stresses after initiation of the heat removal by a cold water from emergency core cooling system. In the paper the conservatism in criteria to avoid brittle ruptures of the fuel elements is discussed. Taking into account the influence of fuel burnup on the property of the cladding and a potential presence of air in the steam, it is believed that criteria of survivability of the zircaloy fuel cladding during LOCA may not be enough conservative.(author)

  20. Review of high burn-up RIA and LOCA database and criteria

    International Nuclear Information System (INIS)

    Vitanza, C.; Hrehor, M.

    2006-01-01

    This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

  1. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M T; Garcia Cuesta, J C; Vallejo Diaz, I; Puebla, Herranz

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  2. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  3. Fission Product Transport Models Adopted in REFPAC Code for LOCA Conditions in PWR and WWER NPPS

    International Nuclear Information System (INIS)

    Strupczewski, A.

    2003-01-01

    The report presents assumptions and physical models used for calculations of fission product releases from nuclear reactors, their behavior inside the containment and leakages to the environment after large break loss of coolant accident LB LOCA. They are the basis of code REFPAC (RElease of Fission Products under Accident Conditions), designed primarily to represent significant physical processes occurring after LB LOCA. The code describes these processes using three different models. Model 1 corresponds to established US and Russian practice, Model 2 includes all conservative assumptions that are in agreement with the actual state-of-the-art, and Model 3 incorporates formulae and parameter values actually used in EU practice. (author)

  4. Best estimate LB LOCA approach based on advanced thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Sauvage, J.Y.; Gandrille, J.L.; Gaurrand, M.; Rochwerger, D.; Thibaudeau, J.; Viloteau, E.

    2004-01-01

    Improvements achieved in thermal-hydraulics with development of Best Estimate computer codes, have led number of Safety Authorities to preconize realistic analyses instead of conservative calculations. The potentiality of a Best Estimate approach for the analysis of LOCAs urged FRAMATOME to early enter into the development with CEA and EDF of the 2nd generation code CATHARE, then of a LBLOCA BE methodology with BWNT following the Code Scaling Applicability and Uncertainty (CSAU) proceeding. CATHARE and TRAC are the basic tools for LOCA studies which will be performed by FRAMATOME according to either a deterministic better estimate (dbe) methodology or a Statistical Best Estimate (SBE) methodology. (author)

  5. Complete BWR--EM LOCA analysis using the WRAP--EM system

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Gregory, M.V.; Buckner, M.R.

    1979-01-01

    The Water Reactor Analysis Package, Evaluation Model (WRAP--EM), provides a complete analysis of postulated loss-of-coolant accidents (LOCA's) in light--water nuclear power reactors. The system is being developed at the Savannah River Laboratory (SRL) for use by the Nuclear Regulatory Commission (NRC) to interpret and evaluate reactor vendor, evaluation model (EM) analyses. The initial version of the WRAP--EM system for analysis of boiling water reactors (BWR's) is operational. To demonstrate the complete capability of the WRAP--BWR--EM system, a LOCA analysis has been performed for the Hope Creek Plant

  6. A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

    The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

  7. Iris small break loca phenomena identification and ranking table (PIRT)

    International Nuclear Information System (INIS)

    Larson, T.K.; Moody, F.J.; Wilson, G.E.; Brown, W.L.; Frepoli, C.; Hartz, J.; Woods, B.G.; Oriani, L.

    2007-01-01

    The international reactor innovative and secure (IRIS) is a modular pressurized water reactor with an integral configuration (all primary system components - reactor core, internals, pumps, steam generators, pressurizer, and control rod drive mechanisms - are inside the reactor vessel). The IRIS plant conceptual design was completed in 2001 and the preliminary design is currently underway. The pre-application licensing process with the United States Nuclear Regulatory Commission (USNRC) started in October 2002. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. If it is not possible to eliminate certain accidents altogether, then the design inherently reduces their consequences and/or decreases their probability of occurring. One of the most obvious advantages of the IRIS Safety-by-Design TM approach is the elimination of large break loss-of-coolant accidents (LBLOCAs), since no large primary penetrations of the reactor vessel or large loop piping exist. While the IRIS Safety-by-Design TM approach is a logical step in the effort to produce advanced reactors, the desired advances in safety must still be demonstrated in the licensing arena. With the elimination of LBLOCA, an important next consideration is to show the IRIS design fulfills the promise of increased safety also for small break LOCAs (SBLOCAs). Accordingly, the SBLOCA phenomena identification and ranking table (PIRT) project was established. The primary objective of the IRIS SBLOCA PIRT project was to identify the relative importance of phenomena in the IRIS response to SBLOCAs. This relative importance, coupled with the current relative state of knowledge for the phenomena, provides a framework for the planning of the continued experimental and analytical efforts. To satisfy the SBLOCA PIRT project objectives, Westinghouse organized an expert panel whose members were carefully selected to insure that the PIRT results reflect internationally

  8. Modeling study of droplet behavior during blowdown period of large break LOCA based on experimental data

    International Nuclear Information System (INIS)

    Sakaba, Hiroshi; Umezawa, Shigemitsu; Teramae, Tetsuya; Furukawa, Yuji

    2004-01-01

    During LOCA (Loss Of Coolant Accident) in PWR, droplets behavior during blowdown period is one of the important phenomena. For example, the spattering from falling liquid film that flows from upper plenum generates those droplets in core region. The behavior of droplets in such flow has strong effect for cladding temperature behavior because these droplets are able to remove heat from a reactor core by its direct contact on fuel rods and its evaporation at the surface. For safety analysis of LOCA in PWR, it is necessary to evaluate droplet diameter precisely in order to predict fuel cladding temperature changing by the calculation code. Based on the test results, a new droplet behavior model was developed for the MCOBRA/TRC code that predicts the droplet behavior during such LOCA events. Furthermore, the verification calculations that simulated some blowdown tests were performed using by the MCOBRA/TRAC code. These results indicated the validity of this droplet model during blow down cooling period. The experiment was focused on investigating the Weber number of steady droplet in the blow down phenomenon of large break LOCA. (author)

  9. Core heatup prediction during SB LOCA with RELAP5/MOD3.2.2 Gamma

    International Nuclear Information System (INIS)

    Parzer, I.; Mavko, B.; Petelin, S.

    2001-01-01

    The paper focuses on the phenomena leading to core uncovering and heatup during the SB LOCA and the ability of RELAP5/MOD3.2.2 Gamma to predict core overheating. The code prediction has been compared to the three experiments, one conducted on the separate effect test facility NEPTUN in Switzerland and the other two conducted on two integral test facilities, PMK-2 in Hungary and PACTEL facility in Finland. In the case of a series of boiloff experiments performed on the NEPTUN test facility the influence of the two correlations available in MOD3.2.2 Gamma for determining interphase drag has been studied. In the case of IAEA-SPE-4 experiment simulation on PMK-2 facility the main goal of the analysis was to study the adequate modeling of the hexagonal core channel with 19-rod bundle and the phenomena during the core uncovering. The third analyzed experiment, OECD-ISP-33, was performed on PACTEL facility to study different natural circulation modes during SB LOCA. The analysis also focused on the final stage of this SB LOCA experiment, when core dryout and heatup was observed due to gradual emptying of the primary system. Following the experience the appropriate modeling options have been used to achieve better representation of the important phenomena during the SB LOCA.(author)

  10. Special LOFT features for improved monitoring and survival of LOCA transients

    International Nuclear Information System (INIS)

    Goodrich, L.D.; Leach, L.P.; Klingler, T.B.; Morrow, J.C.; Phoenix, W.C.; Satterwhite, D.G.; Sumpter, K.C.; Rouhani, S.Z.; Welland, H.J.

    1980-01-01

    LOFT is designed to monitor and survive Loss-Of-Coolant-Accidents (LOCAs). This report presents the primary design difference from LPWRs that were required to accomplish this. These design differences may be of interest to the nuclear power generator industry. This report should be revised semi-annually or as developments in the LOFT Program require

  11. Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code

    International Nuclear Information System (INIS)

    Perianez Alvarez, V.

    2013-01-01

    The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

  12. Methodology for LOCA analysis and its qualification procedures for PWR reload licensing

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1986-01-01

    The methodology for LOCA analysis developed by FURNAS and its qualification procedure for PWR reload licensing are presented. Digital computer codes developed by NRC and published collectively as the WREM package were modified to get versions that comply to each requirement of Brazilian Licensing Criteria. This metodology is applied to Angra-1 basic case to conclude the qualification process. (Author) [pt

  13. Fitness for service after a LOCA: A process applied to Pickering NGS Unit 2

    International Nuclear Information System (INIS)

    McLean, J.A.; Beaton, D.L.

    1996-01-01

    The fitness for service process provides a unique proven methodology for assessing and correcting post-LOCA damage, essential to plant restart. The process uses the as-built plant configuration for modelling input and features self correcting feedback from inspection to validate assessment models. This paper focuses on the process steps and the infrastructure necessary to execute the process

  14. Safety assessment of the advanced CANDU reactor in postulated LOCA/LOECC events

    International Nuclear Information System (INIS)

    Hazen Hezhi Fan; Zoran Bilanovic

    2005-01-01

    The Advanced CANDU Reactor TM (ACR TM ) retains the proven strengths and features of CANDU reactors, and incorporates innovative new features and state-of-the-art technology. In addition to the enhanced emergency core cooling system, the reserve water system is designed to be available to inject reserve water by gravity into the reactor inlet headers after a postulated loss-of-coolant accident (LOCA). To assist in the ACR design and analysis of beyond the design basis events, simulations are needed to demonstrate the effectiveness of these two independent systems on core cooling, and to assess the consequences of the postulated accident coincident with the impairment of either of the two systems. The current paper is subject to an assessment of a postulated large LOCA coincident with loss of the emergency core cooling (LOECC) system. A postulated LOCA/LOECC has very low probability, in the range usually associated with severe core damage events. However, in the CANDU design, including ACR, the presence of moderator water surrounding the fuel channels acts as an effective heat sink, together with other safety features, to prevents severe core damage following a postulated LOCA/LOECC. Therefore, it is possible to analyse LOCA/LOECC using the same deterministic tools that are used for analysis of events with much higher frequencies, in the design basis event range. The assessment is conducted based on the current ACR-700 design. However, the analysis methodology, scope, computer tools, and the results in principle, are applicable to larger ACR designs. This assessment includes system (circuit), fuel channel, and fuel analyses. Some assessment results are needed in subsequent moderator analysis and containment analysis. In the assessment, several simulations were performed to analyse the full circuit and individual fuel channel transient behaviours, as well as the fission product release behaviour. The assessment has captured the key responses of the reactor heat

  15. Global vibrations in the wetwell condensation process caused by LOCA in BWR plants

    International Nuclear Information System (INIS)

    Bjoerndahl, O.; Andersson, Magnus

    1998-12-01

    During the last years a substantial part of third part review work related to dynamical loadings has been review of loading specifications dealing with vibrations in containment building related to so called LOCA-events in Swedish BWR plants. Compared to other loading categories characterised as global vibrations these secondary effects of LOCA-events are complex to analyse. One experience from the review work at SAQ up to now is that it is not fully clear what prediction methods and what model idealisations are the most adequate for structural integrity verification on mechanical systems as pressure vessels and piping under such loading conditions. At SAQ Teknik a project work has been carried out to investigate the general status of the methodology used today in Sweden and a work to in the long term develop simplified prediction models and methods for the loading categories condensation oscillations (CO) and chugging (CH). The work was initially concentrated on a study of the background of the methodology which was developed for these type of loading in American BWR-containments of the Mark-II design. The methodology was developed by General Electric, GE, in cooperation with the Mark-II plant owners. The methodology used in Sweden to predict vibrations in BWR containments of this design is with some minor modifications very close to technique developed by GE. The methodology developed by GE is the only accepted by USNRC for the Mark-II design and could be found as reference in Standard Review Plan 6.2.1.1.C, Rev 6 - August 1984. Based on identical physical assumptions about the dynamic behaviour of the building structure and the water in the suppression pool, mathematical models are derived in this report for predictions of secondary structure response spectra for loading conditions as global vibrations during CO and CH. Based on parameters identified by so called one pipe experiments responses my be predicted. By use of these derived mathematical models as a

  16. Identification of Error of Commissions in the LOCA Using the CESA Method

    Energy Technology Data Exchange (ETDEWEB)

    Tukhbyet-olla, Myeruyert; Kang, Sunkoo; Kim, Jonghyun [KEPCO international nuclear graduate school, Ulsan (Korea, Republic of)

    2015-10-15

    An Errors of commission (EOCs) can be defined as the performance of any inappropriate action that aggravates the situation. The primary focus in current PSA is placed on those sequences of hardware failures and/or EOOs that lead to unsafe system states. Although EOCs can be treated when identified, a systematic and comprehensive treatment of EOC opportunities remains outside the scope of PSAs. However, some past experiences in the nuclear industry show that EOCs have contributed to severe accidents. Some recent and emerging human reliability analysis (HRA) methods suggest approaches to identify and quantify EOCs, such as ATHEANA, MERMOS, GRS, MDTA, and CESA. The CESA method, developed by the Risk and Human Reliability Group at the Paul Scherrer Institute, is to identify potentially risk-significant EOCs, given an existing PSA. The main idea underlying the method is to catalog the key actions that are required in the procedural response to plant events and to identify specific scenarios in which these candidate actions could erroneously appear to be required. This paper aims at identifying EOCs in the LOCA by using the CESA method. This study is focused on the identification of EOCs, while the quantification of EOCs is out of scope. Then, this paper applies the CESA method to the emergency operating procedure (EOP) of LOCA for APR1400. Finally, this study presents potential EOCs that may lead to the aggravation in the mitigation of LOCA. This study has identified the EOC events for APR1400 in the LOCA using CESA method. The result identified three candidate EOCs event using operator action catalog and RAW cutset of LOCA. These candidate EOC events are inappropriate terminations of safety injection system, safety injection tank and containment spray system. Then after reviewing top 100 accident sequences of PSA, this study finally identified one EOC scenario and EOC path, that is, inappropriate termination of safety injection system.

  17. The large break LOCA evaluation method with the simplified statistic approach

    International Nuclear Information System (INIS)

    Kamata, Shinya; Kubo, Kazuo

    2004-01-01

    USNRC published the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology to large break LOCA which supported the revised rule for Emergency Core Cooling System performance in 1989. In USNRC regulatory guide 1.157, it is required that the peak cladding temperature (PCT) cannot exceed 2200deg F with high probability 95th percentile. In recent years, overseas countries have developed statistical methodology and best estimate code with the model which can provide more realistic simulation for the phenomena based on the CSAU evaluation methodology. In order to calculate PCT probability distribution by Monte Carlo trials, there are approaches such as the response surface technique using polynomials, the order statistics method, etc. For the purpose of performing rational statistic analysis, Mitsubishi Heavy Industries, LTD (MHI) tried to develop the statistic LOCA method using the best estimate LOCA code MCOBRA/TRAC and the simplified code HOTSPOT. HOTSPOT is a Monte Carlo heat conduction solver to evaluate the uncertainties of the significant fuel parameters at the PCT positions of the hot rod. The direct uncertainty sensitivity studies can be performed without the response surface because the Monte Carlo simulation for key parameters can be performed in short time using HOTSPOT. With regard to the parameter uncertainties, MHI established the treatment that the bounding conditions are given for LOCA boundary and plant initial conditions, the Monte Carlo simulation using HOTSPOT is applied to the significant fuel parameters. The paper describes the large break LOCA evaluation method with the simplified statistic approach and the results of the application of the method to the representative four-loop nuclear power plant. (author)

  18. CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA

    International Nuclear Information System (INIS)

    Hargroves, D.W.; Metcalfe, L.J.; Cheng, Teh-Chin; Wheat, L.L.; Mings, W.J.

    1991-01-01

    1 - Description of problem or function: CONTEMPT-LT was developed to predict the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided for fan cooler and cooling spray engineered safety systems. One to four compartments can be modeled, and any compartment except the reactor system may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. The user determines the compartments to be used, specifies input mass and energy additions, defines heat structure and leakage systems, and prescribes the time advancement and output control. CONTEMPT-LT/28-H (NESC0433/08) includes also models for hydrogen combustion. 2 - Method of solution: The initial conditions of the containment atmosphere are calculated from input values, and the initial temperature distributions through the containment structures are determined from the steady-state solution of the heat conduction equations. A time advancement proceeds as follows. The input water and energy rates are evaluated at the midpoint of a time interval and added to the containment system. Pressure suppression, spray system effects, and fan cooler effects are calculated using conditions at the beginning of a time-step. Leakage and heat losses or gains, extrapolated from the last time-step, are added to the containment system. Containment volume pressure and temperature are estimated by solving the mass, volume, and energy balance equations. Using these results as boundary conditions, the heat conduction equations

  19. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    International Nuclear Information System (INIS)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

    1998-01-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs

  20. Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

    2018-01-01

    Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

  1. Development and Assessment of the Appendix K Version of RELAP5-3D for LOCA Licensing Analysis

    International Nuclear Information System (INIS)

    Liang, Thomas K.S.; Chang, C.-J.; Hung, H.-J.

    2002-01-01

    In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss-of-coolant accident (LOCA) would limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during a LOCA, it generally takes much more resources to develop. Instead, implementation of evaluation models required by Appendix K of 10CFR50 on an advanced thermal-hydraulic platform such as RELAP5, TRAC, etc., also can gain significant margin for the PCT calculation. Through compliance evaluation against Appendix K of 10CFR50, all of the required evaluation models have been implemented in RELAP5-3D. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effects experiments and eight sets of LOCA integral experiments were adopted. Through the assessments against separate-effects experiments, the success of the code modification in accordance with Appendix K of 10CFR50 was demonstrated. Besides, one set of a typical integral large-break LOCA from Loss-of-Fluid Test Facility experiments (L2-5) has also been applied to preliminarily evaluate the integral performance of the Appendix K version of RELAP5-3D. The PCT predicted by the evaluation models is greater than the one from best-estimate calculation in the whole LOCA history with the conservatism of 150 K, and the measured PCTs of L2-5 are also well bounded by the evaluation model calculation. Another seven sets of integral-effect experiments will be further applied in the next step to ensure the reasonable integral conservatism of the newly developed LOCA licensing analysis code (RELAP5-3DK/INER), which can cover all the phases of both large- and small LOCA in one code

  2. AEEW comments on the NNC/CEGB LOCA code validation report RX 440-A

    International Nuclear Information System (INIS)

    Brittain, I.; Bryce, W.M.; O'Mahoney, R.; Richards, C.G.; Gibson, I.H.; Porter, W.H.L.; Fell, J.

    1984-03-01

    Comments are made on the NNC/CEGB report PWR/RX 440-A, Review of Validation for the ECCS Evaluation Model Codes, by K.T. Routledge et al, 1982. This set out to review methods and models used in the LOCA safety case for Sizewell B. These methods are embodied in the Evaluation Model Computer codes SATAN-VI, WREFLOOD, WFLASH, LOCTA-IV and COCO. The main application of these codes is the determination of peak clad temperature and overall containment pressure. The comments represent the views of a group which has been involved for a number of years in the development and application of Best-Estimate methods for LOCA analysis. It is the judgement of this group that, overall, the EM methods can be used to make an acceptable safety case, but there are a number of points of detail still to be resolved. (U.K.)

  3. Assessment of CONTAIN and MELCOR for performing LOCA and LOVA analyses in ITER

    International Nuclear Information System (INIS)

    Merrill, B.J.; Hagrman, D.L.; Gaeta, M.J.; Petti, D.A.

    1994-09-01

    This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe accident thermal hydraulic analyses. This code will require the least modification to be appropriate for calculating thermal hydraulic behavior in ITER relevant conditions that include vacuum, cryogenics, ITER temperatures, and the presence of a liquid metal test module. The assessment of the aerosol transport models in these codes concludes that several modifications would have to be made to CONTAIN and/or MELCOR to make them applicable to the aerosol transport part of severe accident analysis in ITER

  4. Fuel-rod response during the large-break LOCA Test LOC-6

    International Nuclear Information System (INIS)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% 235 U). Each rod was surrounded by an individual flow shroud

  5. Radial heat transfer from fuel to moderator during LOCAs for CANDU PHW reactors

    International Nuclear Information System (INIS)

    Hildebrandt, J.G.; So, C.B.; Gillespie, G.E.; MacLean, G.

    1983-01-01

    In a postulated CANDU-PHW loss-of-coolant accident (LOCA) with coincident impaired emergency cooling, the axial transport of heat from the fuel by convection is reduced. This reduction in heat removal causes the fuel to heat up and the radial heat transfer to the moderator to become significant. This paper deals with two codes that predict the thermal response of fuel channels under LOCA conditions. New channel thermal radiation models in both RAMA, a thermalhydraulic code, and CHAN II, a fuel channel thermo-chemical code, are presented and their predictions are compared with the experimental results of an electrically heated bundle of 37 fuel pins. A second experiment, involving a single heated pin in a channel with flowing steam, is presented. The predictions of RAMA and CHAN II are compared with this experiment to verify the codes' thermo-chemical models. There is good agreement between the predictions of both codes and the experimental results

  6. Analysis of LOCA/LOECC with a non-stop CATHENA simulation

    International Nuclear Information System (INIS)

    Sabourin, G.; Huynh, H.M.

    1997-01-01

    This paper documents a new approach which simulates without interruption the blowdown and the post-blowdown portions of a LOCA/LOECC. The blowdown portion is simulated first with the pressures, enthalpies, and void fractions of the headers as boundary conditions. The transient inlet header flowrates are written to a file. The blowdown portion is then simulated again with the inlet header flowrates as boundary conditions. At the end of the blowdown, the flowrates are gradually changed to obtain the desired constant gas flowrate of the post-blowdown portion. This new approach was applied with CATHENA MOD3.5a Rev. 0 for a 20% reactor inlet header break coincident with a total loss of emergency core cooling injection. In summary, this paper shows a successful new approach where the blowdown and the post-blowdown portions of a large LOCA coincident with a total loss of emergency core cooling are simulated continuously. (author)

  7. Embrittlement of pre-hydrided Zircaloy-4 by steam oxidation under simulated LOCA transients

    Energy Technology Data Exchange (ETDEWEB)

    Desquines, J., E-mail: jean.desquines@irsn.fr; Drouan, D.; Guilbert, S.; Lacote, P.

    2016-02-15

    During a Loss Of Coolant Accident (LOCA), the mechanical behavior of high temperature steam oxidized fuel rods is an important issue. In this study, as-received and pre-hydrided axial tensile samples were steam oxidized in a vertical furnace and water quenched in order to simulate a LOCA transient. The samples were then subjected to a mechanical test to determine the failure conditions. Two different rupture modes were evidenced; the first one associated to linear elastic fracture mechanics and the second one is associated to sample failure without applied load. The oxidized cladding fracture toughness was determined relying on intensive metallographic analysis. The sample failure conditions were then back predicted confirming that the main rupture parameters are well captured.

  8. Simulation of fuel rod behaviour during various break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Gadalla, A.A.; El-Fawal, M.M.

    1996-01-01

    During loss of coolant accident (LOCAs) course of events, attention focuses on fuel rod cladding temperature behaviour. In this study, the DRUFAN analytical model and LOBI-MOD2 experimental modeling scheme for fuel rod temperature behaviour during C L-Break LOCA in PWRs, are described and discussed. These models are applied for the investigation of fuel rod cladding temperature behaviour during LOCA blowdown phase. A spectrum of selected values representing small, intermediate and large CL- Break sizes are considered in the predictions. The results of the predictions demonstrated that calculated heater rod temperature at steady state as well as the transient period up to 1000 sec are going in good agreement with the measured values. However above 1000 sec the calculated temperatures are higher than the measured values. This indicates that code predictions in this period are conservative. The results indicated also that, in case of small CL-break LOCA (0.01 A and 0.01 and 0.03 A), the heater rod cladding temperature don't rise above saturation temperature. However, on the top of the heater rod, DNB is occurred in case of 0.03 A CL break, while for 0.01 A break, DNB didn't occur. In case of intermediate and large CL-break; (0.05 A, 0.10 A and 1 A), the results showed that, the heater rod cladding temperature exceeded the saturation temperature and DNB prevailed in upper and intermediate sections of the core. 15 figs., 2 tabs

  9. Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

    International Nuclear Information System (INIS)

    Axford, D.J.; Martin, I.C.; McAuley, S.J.

    1981-04-01

    This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

  10. Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lu, S; Streit, R D; Chou, C K

    1980-01-01

    This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10{sup -12}). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

  11. Lumped-parameter modeling of PWR downcomer and pressurizer for LOCA conditions

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Saha, P.; Dubow, A.A.

    1978-01-01

    Two lumped-parameter models, one for a PWR downcomer and the other for a pressurizer, are presented. The models are based on the transient, nonhomogeneous, drift-flux description of two-phase flow, and are suitable for simulating a hypothetical LOCA condition. Effects of thermal nonequilibrium are incorporated in the downcomer model, whereas the pressurizer model can track the interfaces among various flow regimes. Semiimplicit numerical schemes are used for solution. Encouraging results have been obtained for both the models. (author)

  12. Simulating a partial LOCA in a narrow channel using the DSNP simulating system

    International Nuclear Information System (INIS)

    Saphier, D.

    2007-01-01

    A partial LOCA accident in a pool type research reactor was investigated. A new MTR type fuel channel model for the DSNP simulation system was developed; permitting detailed axial and radial temperature distribution. New and older heat transfer correlations were incorporated in the model. Simulation for water levels of 14 and 35 cm in a 62 cm channel were performed. The resulting maximum temperatures remain significantly below the aluminium melting point, and no damage to the core will take place under these conditions

  13. Comparison of LOCA safety analysis in the USA, FRG, and Japan

    International Nuclear Information System (INIS)

    Leach, L.P.; Ybarrondo, L.J.; Hicken, E.F.; Tasaka, K.

    1983-01-01

    The bases for loss-of-coolant accident (LOCA) safety analysis required by reactor licensing regulations in the United States of America (USA), Federal Republic of Germany (FRG), and Japan are investigated and related to new data obtained since the regulations were established. The licensing approaches used in the three countries are similar in that a conservative calculation is called for, the necessary conservatism is unspecified, and new research data have had only limited effect on changing the regulations

  14. PHEBUS program: first results on PWR fuel behaviour in LOCA conditions

    International Nuclear Information System (INIS)

    Del Negro, R.; Reocreux, M.; Pelce, J.; Legrand, B.; Berna, P.

    1982-09-01

    In the first PHEBUS test with pressurized rods some rods burst and clad temperature reached 1100 0 C in the 25 rods bundle. There is now a lot of valuable experimental results and their analysis is in progress. The phase II on fuel behaviour in case of a large LOCA will start at the beginning of 83. The onset of the SFD program is foreseen to take place on the first months of 85

  15. Fluid-structure coupled dynamic response of PWR core barrel during LOCA

    International Nuclear Information System (INIS)

    Lu, M.W.; Zhang, Y.G.; Shi, F.

    1991-01-01

    This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

  16. Large LOCA-earthquake combination probability assessment - Load combination program. Project 1 summary report

    International Nuclear Information System (INIS)

    Lu, S.; Streit, R.D.; Chou, C.K.

    1980-01-01

    This report summarizes work performed for the U.S. Nuclear Regulatory Commission (NRC) by the Load Combination Program at the Lawrence Livermore National Laboratory to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR-1, is used for this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated with a deterministic fracture mechanics model that incorporates stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without an earthquake, is very small (on the order of 10 -12 ). The probability of a leak was found to be several orders of magnitude greater than that of a complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported. (author)

  17. Effective water cooling of very hot surfaces during the LOCA accident.

    Czech Academy of Sciences Publication Activity Database

    Štepánek, J.; Bláha, V.; Dostál, V.; Entler, Slavomír

    2017-01-01

    Roč. 124, November (2017), s. 1211-1214 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : LOCA * Quenching * Divertor cooling * Heat transfer * Rewetting Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617303733

  18. Best estimate modeling of fuel thermomechanical behaviour in WWER 1000 LB LOCA

    International Nuclear Information System (INIS)

    Valach, M.; Klouzal, J.; Zymak, J.; Dostal, M.

    2009-01-01

    The paper summarizes our calculations of the performance of the WWER 1000 NPP fuel rods during postulated LB LOCA. The thermomechanical modeling was performed by FRAPTRAN using the FRACAS-I mechanical model using the boundary conditions calculated by the ATHLET code. The results and their statistical evaluation are presented, the process of the generalization of gained insight into the best-estimate thermal-hydraulic analyses (BE TM) predictions in order to define a generic BE TM methodology is outlined (authors)

  19. LOCA simulation tests in the RD-12 loop with multiple heat channels

    International Nuclear Information System (INIS)

    Ardron, K.H.; McGee, G.R.; Hawley, E.H.

    1985-11-01

    A series of tests has been performed in the RD-12 loop to study the bahaviour of a CANDU-type, primary heat transport system (PHTS) during the blowdown and injection phases of a loss-of-coolant accident (LOCA). Specifically, the tests were used to investigate flow stagnation and refilling of the core following a LOCA. RD-12 is a pressurized water loop with the basic geometry of a CANDU reactor PHTS, but at approximately 1/125 volume scale. The loop consists of U-tube steam generators, pumps, headers, feeders, and heated channels arranged in the symmetrical figure-of-eight configuration of the CANDU PHTS. In the LOCA simulation tests, the loop contained four horizontal heated channels, each containing a seven-element assembly of indirectly heated, fuel-rod simulators. The channels were nominally identical, and were arranged in parallel pairs between the headers in each half-circuit. Tests were carried out using various restricting orifices to represent pipe breaks of different sizes. The break sizes were specifically chosen such that stagnation conditions in the heated channels would be likely to occur. In some tests, the primary pumps were programmed to run down over a 100-s period to simulate a LOCA with simultaneous loss of pump power. Test results showed that, for certain break sizes, periods of low flow occurred in the channels in one half of the loop, leading to flow stratification and sheath temperature excursions. This report reviews the results of two of the tests, and discusses possible mechanisms that may have led to the low channel flow conditions observed in some cases. Plans for future experiments in the larger scale RD-14 facility are outlined. 5 refs

  20. Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.

    1983-01-01

    Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)

  1. A new high temperature deformation model for zircaloy clad ballooning under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Brzoska, B.; Cheliotis, G.; Kunick, A.; Senski, G.

    1977-01-01

    Assuming Zircaloy clad ballooning occurs predominantly by thermal activated secondary creep, generally a power law is applied to describe the creep rate analytically. According to Norton the creep rate is taken as a power function of the cladding hoop stress multiplied by a numerical constant which is determined by the cladding structural properties and a Boltzmann factor including the creep activation energy, the gas constant and the cladding temperature respectively. As is well known, the stress exponent is not a constant value in the total range of LOCA stresses, but increases steadily with stress. This difficulty is avoided by introducing into the Norton law a plastic flow-factor including a limiting stress, which was derived by G. Senski using plastic crack models from Dugdale and Irwin. For LOCA applications the limiting stress is identified with the burst stress, which is experimentally determined. A total number of about 280 directly heated KWU burst tests including two types of experiments: (i) controlled temperature transient tests, (ii) creep rupture tests, are used to fit the burst stress of KWU zircaloy tubes simulating the whole range of LOCA temperatur

  2. Simulated LOCA Test and Characterization Study Related to High Burn-Up Issue

    International Nuclear Information System (INIS)

    Park, D. J.; Jung, Y. I.; Choi, B. K.; Park, S. Y.; Kim, H. G.; Park, J. Y.

    2012-01-01

    For the safety evaluation of fuel cladding during the injection of emergency core coolant, simulated Loss-of-coolant accident (LOCA) test was performed by using Zircaloy-4 fuel cladding samples. Zircaloy-4 tube samples with and without prehydring were oxidized in a steam environment with the test temperature of 1200 .deg. C. Prehydrided cladding was prepared from as-fabricated Zircaloy-4 to study the effects of hydrogen on mechanical properties of cladding during high temperature oxidation and quench conditions. In order to measure the ductility of the tube samples embrittled by quenching water, ring compression test was carried out by using 8 mm ring sample sectioned from oxidized tube sample and microstructural analysis was also performed after simulated LOCA test. The results showed that hydrogen increases oxygen solubility and pickup rate in the beta layer. This reduces ductility of prehydrided fuel cladding compared with as-fabricated cladding. Trend in ductility decrease for prehydrided sample under simulated LOCA condition was very similar with data obtained from tests conducted using irradiated high burn-up fuel claddings

  3. Calculation of Reactivity Build up in KANUPP core in Case of Large Break LOCA

    International Nuclear Information System (INIS)

    Arshad, M. W.

    2012-01-01

    Loss of Coolant Accident (LOCA) in a Pressurized Heavy Water Reactor (PHWR) leads to coolant expulsion in a primary heat transport system resulting in depressurization and possible core voiding. This results in deterioration of cooling conditions in reactor channels and increase in power before reactor shutdown, leading to higher fuel temperatures.The objective of this thesis is to couple Thermal Hydraulics Data for finding status of 2288 fuel bundles having unique coolant density along with continuous changing state of coolant. WIMCER and CITCER are used for the core calculation in case of LOCA and Thermal Hydraulic Data is obtained from the Thermal Hydraulic code TUF (two unequal flows). These codes are coupled with each other in C programming. Due to degradation of coolant in case of LOCA, the power and reactivity start increasing. Near to 5 mk of reactivity the moderator dump start and reactor goes shut down. The result obtained from these code is followed the same trend as shown in KFSAR. (author)

  4. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  5. A simplified time-dependent recovery model as applied to RCP seal LOCAs

    International Nuclear Information System (INIS)

    Kohut, P.; Bozoki, G.; Fitzpatrick, R.

    1991-01-01

    In Westinghouse-designed reactors, the reactor coolant pump (RCP) seals constantly require a modest amount of cooling. This cooling function depends on the service water (SW) system. Upon the loss of the cooling function due to the unavailability of the SW, component cooling water system or electrical power (station blackout), the RCP seals may degrade, resulting in a loss-of-coolant accident (LOCA). Recent studies indicate that the frequency of the loss of SW initiating events is higher than previously thought. This change significantly increases the core damage frequency contribution from RCP seal failure. The most critical/dominant element in the loss of SW events was found to be the SW-induced RCP seal failure. For these potential accident scenarios, there are large uncertainties regarding the actual frequency of RCP seal LOCA, the resulting leakage rate, and time-dependent behavior. The roles of various recovery options based on the time evolution of the seal LOCA have been identified and taken into account in recent NUREG-1150 probabilistic risk assessment PRA analyses. In this paper, a consistent time-dependent recovery model is described that takes into account the effects of various recovery actions based on explicit considerations given to a spectrum of time- and flow-rate dependencies. The model represents a simplified approach but is especially useful when extensive seal leak rate and core uncovery information is unavailable

  6. LOFT/LP-SB-2, Loss of Fluid Test, Small Hot Leg Break LOCA, Delayed Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The third OECD LOFT experiment was conducted on 14 July 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with delayed pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  7. Evaluation of Gap Conductance Approach for Mid-Burnup Fuel LOCA Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, therefore, the applicability of gap conductance approach on the mid-burnup fuel in LOCA analysis was estimated in terms of the comparison of PCT distribution method means the fuel rod uncertainty is taken into account by the combination of overall uncertainty parameters of fuel rod altogether by use of a simple random sampling(SRS) technique. There are many uncertainty parameters of fuel rod that can change the PCT during LOCA analysis, and these have been identified by the authors' previous work already. But, for the 'best-estimate' LOCA safety analysis the methodology that dose not use the overall uncertainty parameters altogether but used the gap conductance uncertainty alone has been developed to simulate the overall fuel rod uncertainty, because it can represent many uncertainty parameters. Based on this approach, uncertainty range of gap conductance was prescribed as 0.67∼1.5 in audit calculation methodology on LBLOCA analysis. This uncertainty was derived from experimental data of fresh or low burnup fuel. Meanwhile, recent research work identify that the currently utilized uncertainty range seems to be not enough to encompass the uncertainty of mid-burnup fuel. Instead it has to be changed to 0.5∼2.4 for the mid-burnup fuel(30 MWd/kgU)

  8. Evaluation of Gap Conductance Approach for Mid-Burnup Fuel LOCA Analysis

    International Nuclear Information System (INIS)

    Lee, Joosuk; Woo, Swengwoong

    2013-01-01

    In this study, therefore, the applicability of gap conductance approach on the mid-burnup fuel in LOCA analysis was estimated in terms of the comparison of PCT distribution method means the fuel rod uncertainty is taken into account by the combination of overall uncertainty parameters of fuel rod altogether by use of a simple random sampling(SRS) technique. There are many uncertainty parameters of fuel rod that can change the PCT during LOCA analysis, and these have been identified by the authors' previous work already. But, for the 'best-estimate' LOCA safety analysis the methodology that dose not use the overall uncertainty parameters altogether but used the gap conductance uncertainty alone has been developed to simulate the overall fuel rod uncertainty, because it can represent many uncertainty parameters. Based on this approach, uncertainty range of gap conductance was prescribed as 0.67∼1.5 in audit calculation methodology on LBLOCA analysis. This uncertainty was derived from experimental data of fresh or low burnup fuel. Meanwhile, recent research work identify that the currently utilized uncertainty range seems to be not enough to encompass the uncertainty of mid-burnup fuel. Instead it has to be changed to 0.5∼2.4 for the mid-burnup fuel(30 MWd/kgU)

  9. Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

    International Nuclear Information System (INIS)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

  10. Statistics and integral experiments in the verification of LOCA calculations models

    International Nuclear Information System (INIS)

    Margolis, S.G.

    1978-01-01

    The LOCA (loss of coolant accident) is a hypothesized, low-probability accident used as a licensing basis for nuclear power plants. Computer codes which have been under development for at least a decade have been the principal tools used to assess the consequences of the hypothesized LOCA. Models exist in two versions. In EM's (Evaluation Models) the basic engineering calculations are constrained by a detailed set of assumptions spelled out in the Code of Federal Regulations (10 CFR 50, Appendix K). In BE Models (Best Estimate Models) the calculations are based on fundamental physical laws and available empirical correlations. Evaluation models are intended to have a pessimistic bias; Best Estimate Models are intended to be unbiased. Because evaluation models play a key role in reactor licensing, they must be conservative. A long-sought objective has been to assess this conservatism by combining Best Estimate Models with statisticallly established error bounds, based on experiment. Within the last few years, an extensive international program of LOCA experiments has been established to provide the needed data. This program has already produced millions of measurements of temperature, density, and flow and millions of more measurements are yet to come

  11. Data management system for full core LOCA-analysis using TRANSURANUS

    International Nuclear Information System (INIS)

    Maertens, D.; Spykman, G.

    2005-01-01

    A data management system has been developed to perform full core pin by pin calculations of normal operation and (LOCA-) transient behaviour of fuel rods. The system automatically generates the input from a data base, controls the fuel rod calculations and provides a powerful tool for visualising the results. The full core pin by pin analysis now allows to use specific power histories, rod geometries and material data as well as enveloping data. Fuel rod code Transuranus is used for the normal operation and the transient phase in one run, thus assuring that the calculated rod properties of the normal operation (pre-transient) phase are handed over in all detail and not compressed to the transient phase. Transuranus has been upgraded with respect to high temperature models for Zry and M5 TM -cladding for creep, oxidation, heat rate dependent phase transition and anisotropy in the α and the mixed crystal phase. Parameter studies have been carried out to investigate the influence of using rod specific power histories instead of enveloping power histories in a full core analysis. The results show a significant increase in the ratio of failed fuel rods during a LOCA transient from 0.12% to approx. 50%. Another study for a typical PWR LOCA transient shows very good correlation between the distribution of failed fuel rods and rods with significant ballooning. (author)

  12. Analysis of a convection loop for GFR post-LOCA decay heat removal

    International Nuclear Information System (INIS)

    Williams, W.C.; Hejzlar, P.; Saha, P.

    2004-01-01

    A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA gas-cooled fast reactor (GFR). The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO 2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO 2 outdoes helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops. (authors)

  13. Long-term aging and loss-of-coolant accident (LOCA) testing of electrical cables

    International Nuclear Information System (INIS)

    Nelson, C.F.; Gauthier, G.; Carlin, F.

    1996-10-01

    Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 conductor with cross-linked polyethylene (XLPE) insulation and a Hypalon jacket. (3) French 3 conductor with EPR insulation and a Hypalon jacket. (4) French coaxial with polyethylene (PE) insulation and a PE jacket. The data represent up to 5 years of simultaneous aging where the cables were exposed to identical aging radiation doses at either 40 degrees C or 70 degrees C; however, the dose rate used for the aging irradiation was varied over a wide range (2-100 Gy/hr). Aging was followed by exposure to simulated French LOCA conditions. Several mechanical, electrical, and physical-chemical condition monitoring techniques were used to investigate the degradation behavior of the cables. All the cables, except for the French PE cable, performed acceptably during the aging and LOCA simulations. In general, cable degradation at a given dose was highest for the lowest dose rate, and the amount of degradation decreased as the dose rate was increased

  14. Prediction of Golden Time for Recovering the Safety Injection System in Severe LOCA Circumstances

    International Nuclear Information System (INIS)

    Yoo, Kwae Hwan; Kim, Dong Young; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun

    2015-01-01

    In this study, the core uncovery and RV failure according to LOCA break sizes were analyzed by using the MAAP4 code when safety injection system (SIS) was not operating normally. We predicted the golden time of SIS recovery for accomplishing the reactor cold shutdown and preventing RV failure. MAAP4 code was used for severe accident analysis. The LOCA simulations were performed with break size in order to predict the golden time to recovery SIS. We predicted the golden time according to the SIS operation cases through the simulation of OPR1000. When LOCA occurred, the normal operation of SIS is very important in maintaining the integrity of NPPs. However if the SIS does not work or its actuation is delayed due to failure of the equipment, the DBA will lead to a severe accident. In this study, accident situations that SIS does not work normally were assumed and a number of MAAP4 code simulations were conducted. In addition, core uncovery time and RV failure time were predicted. If the recovery time of SIS for accident recovery is predicted, the core will not be exposed through appropriate action

  15. LOFT/LP-SB-1, Loss of Fluid Test, Small Hot Leg Break LOCA, Early Pump

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The second OECD LOFT experiment was conducted on 23 June 1983. It simulated a 3-in (7.62 cm) equivalent break diameter located in the hot leg of the operating loop. The major objective of this experiment was to determine system transient characteristics for small hot leg break loss-of-coolant accidents with early pump trip. The experiment was conducted from initial temperature and pressure conditions representative of typical commercial PWRs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  16. Evaluation of control room habitability in case of LOCA for Maanshan NPP using codes RADTRAD, HABIT and ALOHA

    International Nuclear Information System (INIS)

    Hsu, Wen-Sheng; Wang, Jong-Rong; Chen, Hsiung-Chih; Chiang, Yu; Chen, Shao-Wen; Shih, Chunkuan

    2018-01-01

    The method for the evaluation of the control room habitability is presented in this paper with focus on Maanshan PWR nuclear power plant (NPP) using the codes RADTRAD, HABIT, and ALOHA. Therefore, this paper is divided into two parts: The first part is the evaluation of the cumulative dose at the control room, the exclusion area boundary (EAB) and the low population zone (LPZ) in case of an design basis loss of coolant accident (DBA/LOCA). For this first part, the Maanshan NPP models of the code RADTRAD/SNAP were used for the analysis. The second part is the evaluation of the control room habitability under the assumption of CO 2 storage burst. For this part, the HABIT and ALOHA codes were used. As result it was seen that the RADTRAD calculation results are below the failure criteria of standard review plan (SRP) and 10 CFR 100.11. The HABIT and ALOHA results are below the R.G. 1.78 failure criteria. These results indicate that Maanshan NPP' habitability can be maintained under the above conditions.

  17. An estimation of the accident sequence of the LOCA groups for the PSA model of the KSNP

    International Nuclear Information System (INIS)

    Han, Seok Jung; Yang, Joon Eon

    2004-01-01

    A new trend of the probabilistic safety assessment (PSA) technology is to improve and enhance the current PSA model to be adequate for risk-informed applications (RIA). Requirements of a PSA model for the RIA are summarized as (1) reduction of the conservatism in the model utilizing all available information and (2) consideration of the specific features of a plant as designed, as operated. This is because the PSA based on conservatism and insufficient consideration of the plant-specific features resulted in a shadow effect on the assessment results. When a PSA model is used in a risk-informed application, more precise risk-information is more helpful to decision making process, so the reduction of the conservatism and the consideration of the plant-specific features in a PSA model are the most essential elements. Recently, an effort has been performed to modify the current PSA model for the Korea Standard Nuclear Power plant (KSNP) to be used in risk-informed applications. A re-estimation of the accident sequence of the loss of coolant accident (LOCA) groups for the PSA model of the KSNP has been performed

  18. PACTEL ISP-33. MELCOR assessment

    International Nuclear Information System (INIS)

    Siccama, N.B.

    1995-09-01

    The OECD/CSNI International Standard Problem (ISP-33) experiment was a natural-circulation experiment with a stepwide reduced primary coolant inventory in the PACTEL facility. The MELCOR code has been used to simulate this experiment. The main goal of these post-test calculations was to assess MELCOR on one- and two-phase natural-circulation phenomena which occur in Eastern European VVER plants in case of LOCA conditions. A base case and several senstivity calculations have been performed. In addition, the MELCOR results have been compared to results obtained by the RELAP5 code. Different natural-circulation modes have been identified during the experiment and simulated with MELCOR in the analyses of the ISP-33 experiment. These are successively: The single-phase liquid flow, the transient two-phase flow, the steady two-phase flow, and the boiler-condenser heat removal. These regimes, except the transient two-phase flow, are calculated in good agreement with the experiment. Special attention has been paid to the modeling of the two-phase flow in the hot legs of the PACTEL facility. Sensitivity calculation have shown that the results to a large extent are influenced by the nodalization of the hot legs and the opening heights of the hot-leg flow paths. Other senstivity calculations have shown that the time step and the core model do not influence the results, and accurate values for form loss coefficients and properties of the insulation are not necessary. The integrated MELCOR code is not inferior to the mechanistic RELAP5 code for the PACTEL ISP-33 post-test calculations. Some phenomena are modeled even better by MELCOR, because of the ability fit MELCOR parameters. (orig.)

  19. Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

    Energy Technology Data Exchange (ETDEWEB)

    Fleming, K.N., E-mail: KarlFleming@comcast.net [KNF Consulting LLC, Spokane, WA (United States); Lydell, B.O.Y. [SIGMA-PHASE INC., Vail, AZ (United States)

    2016-08-15

    Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

  20. Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

    International Nuclear Information System (INIS)

    Fleming, K.N.; Lydell, B.O.Y.

    2016-01-01

    Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

  1. Probabilistic based design rules for intersystem LOCAS in ABWR piping

    International Nuclear Information System (INIS)

    Ware, A.G.; Wesley, D.A.

    1993-01-01

    A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem loss-of-coolant accident (ISLOCA). The methodology was applied to various sizes of carbon and stainless steel piping designed to advanced boiling water reactor (ABWR) temperatures and pressures

  2. LOFT/L3-6, Loss of Fluid Test, 6. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the sixth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were running. The experiment was conducted on 10 December 1980

  3. LOFT/L3-5, Loss of Fluid Test, 5. NRC L3 Small Break LOCA Experiment

    International Nuclear Information System (INIS)

    1991-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This was the fifth in the NRC L3 Series of small-break LOCA experiments. A 10-cm (2.5-in.) cold-leg non-communicative-break LOCA was simulated. Pumps were shut off. The experiment was conducted on 29 September 1980

  4. Compendium of ECCS [Emergency Core Cooling Systems] research for realistic LOCA [loss-of-coolant accidents] analysis: Final report

    International Nuclear Information System (INIS)

    1988-12-01

    In the United States, Emergency Core Cooling Systems (ECCS) are required for light water reactors (LWRs) to provide cooling of the reactor core in the event of a break or leak in the reactor piping or an inadvertent opening of a valve. These accidents are called loss-of-coolant accidents (LOCA), and they range from small leaks up to a postulated full break of the largest pipe in the reactor cooling system. Federal government regulations provide that LOCA analysis be performed to show that the ECCS will maintain fuel rod cladding temperatures, cladding oxidation, and hydrogen production within certain limits. The NRC and others have completed a large body of research which investigated fuel rod behavior and LOCA/ECCS performance. It is now possible to make a realistic estimate of the ECCS performance during a LOCA and to quantify the uncertainty of this calculation. The purpose of this report is to summarize this research and to serve as a general reference for the extensive research effort that has been performed. The report: (1) summarizes the understanding of LOCA phenomena in 1974; (2) reviews experimental and analytical programs developed to address the phenomena; (3) describes the best-estimate computer codes developed by the NRC; (4) discusses the salient technical aspects of the physical phenomena and our current understanding of them; (5) discusses probabilistic risk assessment results and perspectives, and (6) evaluates the impact of research results on the ECCS regulations. 736 refs., 412 figs., 66 tabs

  5. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  6. High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

    International Nuclear Information System (INIS)

    Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

    2013-01-01

    In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

  7. Prediction of moderator temperature under 35% RIH break LOCA with LOECC in CANDU calandria vessel

    International Nuclear Information System (INIS)

    Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung; Lee, Jae Yung

    2004-01-01

    A CANDU reactor has the unique safety features with the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors such as a PWR. One of the safety features is that the heavy water moderator is continuously cooled, providing with a heat sink for the decay heat produced in the fuel when there is the LOCA with the coincident failure of the emergency coolant injection (ECI) system. Under such a dual failure condition, the hot pressure tube (PT) would deform into contacting with the calandria tube (CT), providing with an effective heat transfer path from the fuel to the moderator. Following PT/CT contact, there is the spike of the heat flux in the moderator surrounding the CT, which could lead to sustained CT dryout. The prevention of the CT dryout depends on available local moderator subcooling. Higher moderator temperature (or lower subcooling) would decrease the margin of the CTs to dryout. As for LOCAs with coincident loss of the ECI, fuel channel integrity depends on the capability of the moderator as an ultimate heat sink. In this regard, the Canadian Nuclear Safety Commission (CNSC) had categorized the temperature prediction for the moderator cooling integrity as a general action item (GAI) and had recommended that a series of experimental works should be performed to verify the evaluation codes comparing with the results of three-dimensional experimental data. However, although a couple of computer codes were used to predict moderator temperature prediction for those problems, they could not be adequately validated due to the uncertainty of temperature prediction. In this work, the temperature prediction under the transient condition of LOCA with loss of emergency core cooling (LOECC) in a CANDU reactor is conducted using the optimized calculation scheme from the previous work

  8. An application of RELAP5/MOD3 to the post-LOCA long term cooling performance evaluation

    International Nuclear Information System (INIS)

    Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

    1998-01-01

    A realistic long-term calculation to be used in the post-LOCA long term cooling (LTC) analysis is described in this study, which was required to resolve the post-LOCA LTC issues including the concern on boric acid precipitation in the reactor core. The analysis scope is defined according to the LTC plan of UCN Units 3/4 and the plant calculation model are developed suitable to the LTC procedure. The LTC sequences following the cold leg small break LOCAs of 0.02 ft2 to 0.5 ft2 are calculated by RELAP5/ MOD3.2.2. Based on the calculation results, the establishment of shutdown cooling system entry condition and the behavior of boron transport are evaluated. The effect of model simplification is also investigated

  9. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    International Nuclear Information System (INIS)

    Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

    2015-01-01

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  10. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of); Chang, Soon Heung [Handong Global University, 558, Handong-ro, Buk-gu, Pohang Gyeongbuk 37554 (Korea, Republic of); Choi, Yu Jung [Korea Hydro and Nuclear Power Co.—Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 34101 (Korea, Republic of); Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of)

    2015-12-15

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  11. Computation programs for the thermofluidodynamic transient analysis in the containment system following a LOCA

    International Nuclear Information System (INIS)

    Gorlandi, A.; Mazzini, M.; Oriolo, F.

    1979-01-01

    This works briefly describes the features of the computation codes available at the Istituto di Impianti Nucleari of the Pisa University for the analysis of the thermofluidodynamic transient in the containment system of a nuclear power plant following a LOCA (RELAP 4/MOD.S, COMPARE, FUMO and CONTEMPT-LT/026). More details are contained in the Annex. Particular attention has been devoted to the opportunity to study, through the computation codes, the effects of the sub division of a full pressure containment system

  12. Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients

    International Nuclear Information System (INIS)

    Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

    1980-01-01

    The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations

  13. LDA measurement of droplet behavior across tie plate during dispersed flow portion of loca reflood

    International Nuclear Information System (INIS)

    Lee, S.L.; Srinivasan, J.; Cho, S.K.

    1980-01-01

    The flow of an air-water droplet dispersion in a simulated 3-D test section in the reflood portion of LOCA was studied. For this purpose, a new scheme of Laser-Doppler Anemometry for the simultaneous measurement of size and velocity of large-size [0.5 mm-6 mm] droplets was developed and utilized. It was observed that the size distribution of the reentrained droplets depends mainly on the flow regimes and is essentially independent of that of the incoming dispersion below the tie plate. 8 refs

  14. Evaluation of simulated-LOCA tests that produced large fuel cladding ballooning

    International Nuclear Information System (INIS)

    Powers, D.A.; Meyer, R.O.

    1979-02-01

    A description is given of the NRC review and evaluation of simulated-LOCA tests that produced large axially extended ballooing in Zircaloy fuel cladding. Technical summaries are presented on the likelihood of the transient that was used in the tests, the effects of temperature variations on strain localization, and the results of other similar experiments. It is concluded that (a) the large axially extended deformations were an artifact of the experimental technique, (b) current NRC licensing positions are not invalidated by this new information, and (c) no new research programs are needed to study this phenomenon

  15. Reflooding phase of the LOCA - state of the art II. Rewetting and liquid entrainment

    International Nuclear Information System (INIS)

    Elias, E.; Yadigaroglu, G.

    1977-01-01

    Understanding the mechanisms by which hot fuel rods quench and the physics of liquid droplet entrainment is important for the analysis of the reflooding phase of the LOCA. Published models of the rewetting process include simple one-dimensional solutions. The basic physical assumptions of these models and the numerical values assigned to the various parameters, as well as empirical rewetting correlations are discussed. The various mechanisms for liquid droplet entrainment and analytical formulations of the critical gas velocity and of the droplet diameter at the onset of entrainment are reviewed

  16. Experimental studies on mitigation of LOCA for a high flux research reactor

    International Nuclear Information System (INIS)

    Saxena, A.K.

    2006-01-01

    Experimental studies on the rewetting behaviour of hot vertical annular channels were performed to study the mitigation of consequences of loss of coolant accident (LOCA) for a high flux research reactor. Studies were carried out to study the rewetting behaviour with hot inner tube, for bottom flooding and top flow rewetting conditions. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 litres per minute (l pm) (11.7 x 10 -5 m 3 s -1 ). The initial surface temperature of the inner tube was varied from 200 to 500 degC. (author)

  17. Modelling of WWER fuel rod during LOCA conditions using FEM code ANSYS

    International Nuclear Information System (INIS)

    Bogatyr, S. M.; Krupkin, A. V.; Kuznetsov, V. I.; Novikov, V. V.; Petrov, O. M.; Shestopalov, A. A.

    2013-01-01

    The report presents the results of the computer simulation of the IFA-650.6 experiment, the sixth test in Halden LOCA test project series, performed in May 18, 2007 with a pre-irradiated WWER-440 fuel with maximum burnup of 56 MWd/kgU. The thermo-mechanical analysis was fulfilled with the license finite element ANSYS code package.The calculation was carried out with the 2D axisymmetric and 3D problem definitions. Analysis of the calculational results shows that the ANSYS code can adequately simulate thermo-mechanical behavior of cladding under IFA-650.6 test conditions. (authors)

  18. Investigation on the Behavior of the Jointing Clamps to the Simulated Environmental - LOCA

    International Nuclear Information System (INIS)

    Ivan, P.; Segarceanu, D.; Geambasu, C.

    2002-01-01

    The paper presents the main aspect concerning the electric parameter variation of jointing clamps operating under specific environmental conditions determinate by pressure, temperature and humidity. The testing of jointing clamps capability to meet and exceed the required performances all along its operating life implies the performing of LOCA simulation conditions while the jointing clamps is bright in a relatively short time under conditions equivalent to those at the end of its service life. The paper describes ageing and measurement techniques and the analyses of electric parameter behaviour in these environmental simulated conditions. (author)

  19. R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron Simon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tu, Lei [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize

  20. Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR (PASS)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2015-10-15

    Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability (T < 1300 °C) with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors (SMRs) (NuScale, SMART, IRIS) and a commercial pressurized water reactor (PWR) are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR (PASS), which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident (LOCA) by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—(1) FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, (2) computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and (3) multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of

  1. In-core LOCA-s: analytical solution for the delayed mixing model for moderator poison concentration

    International Nuclear Information System (INIS)

    Firla, A.P.

    1995-01-01

    Solutions to dynamic moderator poison concentration model with delayed mixing under single pressure tube / calandria tube rupture scenario are discussed. Such a model is described by a delay differential equation, and for such equations the standard ways of solution are not directly applicable. In the paper an exact, direct time-domain analytical solution to the delayed mixing model is presented and discussed. The obtained solution has a 'marching' form and is easy to calculate numerically. Results of the numerical calculations based on the analytical solution indicate that for the expected range of mixing times the existing uniform mixing model is a good representation of the moderator poison mixing process for single PT/CT breaks. However, for postulated multi-pipe breaks ( which is very unlikely to occur ) the uniform mixing model is not adequate any more; at the same time an 'approximate' solution based on Laplace transform significantly overpredicts the rate of poison concentration decrease, resulting in excessive increase in the moderator dilution factor. In this situation the true, analytical solution must be used. The analytical solution presented in the paper may also serve as a bench-mark test for the accuracy of the existing poison mixing models. Moreover, because of the existing oscillatory tendency of the solution, special care must be taken in using delay differential models in other applications. (author). 3 refs., 3 tabs., 8 figs

  2. An outcome of nuclear safety research in JAERI. Case study for LOCA, FP, criticality and reprocessing

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Ito, Keishiro; Kawashima, Kei; Katsuki, Chisato; Shirabe, Masashi

    2009-09-01

    An outcome of nuclear safety research done by JAERI was case studied by the bibliometric method. (1) For LOCA (loss-of-coolant accident) a domestic share of JAERI in monoclinic research paper was 63% at the past (20) 1978-1982 but was decreased to 40% at the present 1998-2002. For co-authored papers a domestic share between JAERI and PS (public sectors) is almost zero at past (20) but increased to 4% at the present. Research cooperation is active between Tokyo University and JAERI or between JAERI and Nagoya University. (2) Project-type research is to have a large monopolization in papers and that of basic-type research is to have a large development of research networking (DRN). (3) For FP, a share of co-authored paper is high due to an enhanced cooperation among JAERI-PO (Public Organization)-PS. For criticality, research activity was enhanced after JCO accident, especially at NUCEF. (4) For reprocessing, PS had a monopolistic position with a domestic share of 71% and a share of JAERI was about 20%. (5) LOCA and RIA outputs born by NSR-JAERI coincided partly to those of the Safety Licensing Guidelines but a share of contribution done by JAERI was ambiguous due to the lack of necessary information. (author)

  3. Core heat transfer analysis during a BWR LOCA simulation experiment at ROSA-III

    International Nuclear Information System (INIS)

    Yonomoto, T.; Koizumi, Y.; Tasaka, K.

    1987-01-01

    The ROSA-III test facility is a 1/424-th volumetrically scaled BWR/6 simulator with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA). Heat transfer analyses for 5, 15, 50 and 200% break tests were conducted to understand the basic heat transfer behavior in the core under BWR LOCA conditions and to obtain a data base of post-critical heat flux (CHF) heat transfer coefficients and quench temperature. The results show that the convective heat transfer coefficient of dried-out rods at the core midplane during a steam cooling period is less than approximately 120 W/m 2 K. It is larger than existing data measured at lower pressures during a spray cooling period. Bottom-up quench temperatures are given by a simple equations: The sum of the saturation temperature and a constant of 262 K. Then the heat transfer model in the RELAP4/MOD6/U4/J3 code was revised using the present results. The rod surface temperature behavior in the 200% break test was calculated better by using the revised model although the model is very simple. (orig.)

  4. Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

    International Nuclear Information System (INIS)

    Vigil, R.A.; Jacobus, M.J.

    1994-04-01

    Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent

  5. Mixing phenomena of interest to boron dilution during small break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.; Cheng, Z.

    1995-01-01

    This paper presents the results of a study of mixing phenomena related to boron dilution during small break loss of coolant accidents (LOCAs)in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler-condenser mode. A problem may occur when subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core are examined in this paper. Bounding calculations for boron concentration of coolant entering the core during a small break LOCA in a typical Westinghouse-designed four-loop plant are also presented

  6. Water volume available for ECCS sump recirculation mode following a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Riekert, T. [TUV NORD SysTec (Germany); Rebohm, H. [TUV NORD EnSys Hannover (Germany); Huber, J. [TUV SUD IS (Germany); Brandes, F. [TUV SUD ET (Germany)

    2006-07-01

    In this paper we describe the reviews performed in Germany on the water level in the containment sump after a LOCA and the derived actions. Our view on the issue is from the perspective of the independent safety experts - i.e. TUV SUD Industrie Service (TUV SUD IS), TUV SUD Energietechnik GmbH Baden-Wuerttemberg (TUV SUD ET), TUV NORD EnSys Hannover and TUV NORD SysTec -, which reviewed the analyses of the utilities on behalf of the responsible supervising authorities. Between these expert organizations information were exchanged via the steering committee on nuclear technology of the association of the TUVs (VdTUV). In our paper we describe the analyses on the two safety issues relevant in the connection with the water level in the containment sump: the necessary minimum coverage of suction pipes to avoid inadmissible entrainment of air and the water retention inside the containment after a LOCA. Our description concentrates on PWRs because of the more complex conditions in comparison to BWRs. In conclusion it can be stated that due to the thorough evaluation of operating experience, optimization measures could be derived. In addition, the analyses served the purpose of know-how maintenance. (authors)

  7. Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

    International Nuclear Information System (INIS)

    Ferreira Guimaraes, Antonio Cesar; Franklin Lapa, Celso Marcelo; Lamego Simoes Filho, Francisco Fernando; Cabral, Denise Cunha

    2011-01-01

    Research highlights: → This article presents an uncertainty modelling study using a fuzzy approach. → The AP1000 Westinghouse NPP was used and it is provided of passive safety systems. → The use of advanced passive safety systems in NPP has limited operational experience. → Failure rates and basic events probabilities used on the fault tree analysis. → Fuzzy uncertainty approach was employed to reliability of the AP1000 large LOCA. - Abstract: This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

  8. LOCA assessment experiments in a full-elevation, CANDU-typical test facility

    International Nuclear Information System (INIS)

    Ingham, P.J.; McGee, G.R.; Krishnan, V.S.

    1990-01-01

    The RD-14 thermal-hydraulics test facility, located at the Whiteshell Nuclear Research Establishment, is a full-elevation model representative of a CANDU primary heat transport system. The facility is scaled to accommodate a single, full-scale (5.0 MW, 21 kg/s), electrically heated channel per pass. The steam generators, pumps, headers, feeders and heated channels are arranged in a typical CANDU figure-of-eight geometry. The loop has an emergency coolant injection system (ECI) that may be operated in several modes, including typical features of the various ECI systems found in CANDU reactors. A series of experiments has been performed in RD-14 to investigate the thermal-hydraulic behaviour during the blowdown and injection phases of a loss-of-coolant accident (LOCA). The tests were designed to cover a full range of break sizes from feeder-sized breaks to guillotine breaks in either an inlet or an outlet header. Breaks resulting in channel flow stagnation were also investigated. This paper reviews the results of some of the LOCA tests carried out in RD-14, and discusses some of the behaviour observed. Plans for future experiments in a multiple-channel RD-14 facility, modified to contain multiple flow channels, are outlined. (orig.)

  9. Prototype application of best estimate and uncertainty safety analysis methodology to large LOCA analysis

    International Nuclear Information System (INIS)

    Luxat, J.C.; Huget, R.G.

    2001-01-01

    Development of a methodology to perform best estimate and uncertainty nuclear safety analysis has been underway at Ontario Power Generation for the past two and one half years. A key driver for the methodology development, and one of the major challenges faced, is the need to re-establish demonstrated safety margins that have progressively been undermined through excessive and compounding conservatism in deterministic analyses. The major focus of the prototyping applications was to quantify the safety margins that exist at the probable range of high power operating conditions, rather than the highly improbable operating states associated with Limit of the Envelope (LOE) assumptions. In LOE, all parameters of significance to the consequences of a postulated accident are assumed to simultaneously deviate to their limiting values. Another equally important objective of the prototyping was to demonstrate the feasibility of conducting safety analysis as an incremental analysis activity, as opposed to a major re-analysis activity. The prototype analysis solely employed prior analyses of Bruce B large break LOCA events - no new computer simulations were undertaken. This is a significant and novel feature of the prototyping work. This methodology framework has been applied to a postulated large break LOCA in a Bruce generating unit on a prototype basis. This paper presents results of the application. (author)

  10. Analysis of the bubble condenser structure of WWER-440 NPP under LOCA loading

    International Nuclear Information System (INIS)

    Zeman, P.

    2003-01-01

    Two problems may arise in relation to the title topic: (1) problem with the uplift of the beams I 600 of the first floor, and (2) possible plastic collapse of the wall on the 12th floor. The problems were tacked by computer calculations. The FEM model of the bubble condenser was created in the ANSYS 6.0 environment and analyzed for the pressure loading defined for the LOCA accident in IAEA TECDOC 803. The model of the bubble condenser structure so created included all geometrical and material non-linearities. The duration of the pressure wave was 0.4 s, amplitude 30 kPa. The analyses revealed that a plastic collapse of the tank wall is not the most critical failure mode. Instead, weld connections appear to be the most critical parts of structure. The tank walls are very ductile and the results of the analyses are in agreement with the test simulating the LOCA accident. The tank walls suffered no damage during the tests

  11. LOCA Analysis of KAIST-Micro Modular Reactor with Modified GAMMA+ code

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Bong Seong; Ahn, Yoon Han; Kim, Seong Gu; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The supercritical carbon dioxide (S-CO{sub 2}) power cycle is being seriously investigated around the world due to its simple layout, quite high efficiency around 500 .deg. C turbine inlet temperature, etc. By combining these two ideas, the KAIST research team developed a S-CO{sub 2} cooled SMR, called KAIST-Micro Modular reactor (MMR), which is targeting transportability and electricity supply for remote region. Therefore, requirements of MMR design are factory fabrication of the total system including power conversion system to be transported and air cooling to be independent from the site selection. Until now, steady performances and sizes of components were evaluated. Thus, in this paper a transient performance of the MMR are simulated with special focus on the loss of coolant accident (LOCA) at cold leg pipe. The MMR is a newly suggested innovative small modular reactor concept by the KAIST research team. Since the MMR is cooled by supercritical CO{sub 2}, general safety codes for conventional reactors have limitations. Thus, GAMMA+ code for the transient analysis of a gas-cooled reactor was selected and modified for the S-CO{sub 2} power system. After the modification of GAMMA+ code, LOCA is simulated, which is considered as one of the most limiting accidents in terms of safety of nuclear power plant.

  12. An investigation into fuel pulverization with specific reference to high burn-up LOCA

    International Nuclear Information System (INIS)

    Yagnik, Suresh; Turnbull, James; Noirot, Jean; Walker, Clive; Hallstadius, Lars; Waeckel, N.; Blanpain, P.

    2014-01-01

    To investigate the phenomenon of high burn-up fuel pellet material potentially disintegrating into powder under a rapid temperature transient, such as in a LOCA-type accident scenario, two independent scoping studies were commissioned. The first was to investigate the effect of hydrostatic restraint pressure on Fission Gas Release (FGR) from small samples of highly irradiated fuel (71 MWd/kgU) during a series of rapid temperature ramps. Experimentally, when the FGR increased rapidly during the temperature transients, the fuel was assumed to be 'pulverized', i.e., fragmented into powder. In the second series of experiments, laser heating of small samples was used to investigate the temperature at which fuel pulverization was initiated. Subsequent to fuel disintegration, there was always a spectrum of particle sizes present. The significance of this observation was recognized in the context of extended burn-up operation in commercial reactors. Based on the observation from these investigations, a fuel fragmentation threshold has been discussed and developed. We conclude that fuel disintegration could be of potential importance in limiting the performance and productive lifetime of nuclear fuel. However, since only fuel closely adjacent to ballooned or ruptured cladding would be released in a LOCA-type transient, expulsion of pulverized fuel from the ruptured fuel rod is not considered a safety issue; cooling of the defected assembly remains possible and there is no issue with respect to local criticality. (author)

  13. Severe damage analysis of VVER 1000 following large break LOCA using Astec code

    International Nuclear Information System (INIS)

    Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.; Ghosh, A.K.; Kushwaha, H.S.

    2007-01-01

    Severe accident analysis of a reactor is an important aspect in the evaluation of source term. This in turn helps in emergency planning. An analysis has been carried out for VVER-1000 (V320) reactor following Large Break LOCA (loss of coolant accident) along with Station Blackout (SBO). Computer code ASTEC (jointly developed by IRSN, France, and GRS, Germany) is used for analyzing the transient. This integral code has been designed to be used as reference code for PSA2 studies. Severe accident analysis is carried out for an accident initiated by Large break LOCA along with SBO. Two cases have been analysed with the version ASTEC V1.2-rev1. In the first case hydro-accumulators are considered not available while the second case has been analysed with hydro accumulators. In this paper, ASTEC predictions have been studied for the in-vessel phase of the accident till vessel failure. The vessel failure was observed at 6979 s when accumulators were assumed not available. The vessel failure was quite delayed (19294 s) with operating accumulators. The hydrogen production was found to be very large (22% of total Zr inventory) in the case with accumulators compared to the case without accumulators (1.5% of total Zr inventory)

  14. The phenomenology of a small break LOCA in a complex thermal hydraulic loop

    International Nuclear Information System (INIS)

    Di Marzo, M.; Almenas, K.K.; Hsu, Y.Y.; Wang, Z.

    1988-01-01

    A phenomenological description of the thermal hydraulics events that take place during a simulated Small Break Loss of Coolant Accident (SB-LOCA) is presented. The SB-LOCA transient is described in detail and the various mass and energy transport modes are identified. Similar behavior is observed in other facilities designed for the simulation of this type of accidents. Previous investigations suggest a simple modelling of the phenomena based on fluid mechanic considerations. An extensive experimental program conducted at the experimental facility of the University of Maryland reveals that condensation is a dominant driving force for this type of transients. This finding has significant implications in the modelling of enthalpy transport for some of the flow modes which occur during the transient. In particular it affects the Interruption and Resumption Mode (IRM) during which enthalpy is transported by periodic flow of a two phase mixture. The efforts to predict the flow interruption based on fluid mechanic criteria of phase separation in the hot leg are shown to be misdirected since thermodynamic phenomena taking place in the horizontal portion of the cold legs and in the reactor vessel downcomer are mostly responsible for that transition. For flow resumption to occur the liquid-vapor mixture swelling in the vertical portion of the hot leg determines the occurrence of the liquid spill over the top of the candy cane. (orig.)

  15. Experimental results of the effective water head in downcomer during reflood phase of a PWR LOCA

    International Nuclear Information System (INIS)

    Sudo, Yukio; Murao, Yoshio; Akimoto, Hajime

    1980-08-01

    The results and analysis of an experiment for the effective water head in downcomer with 50mm gap size are described. The main objective of the experiment was to clarify the effect of gap size on reflooding in a PWR LOCA. The effective water head in downcomer is the driving force for feeding emergency coolant into the core during reflood phase of a PWR LOCA. Discussions presented here follow those of a previous report in which experimental results and analysis were described for the case of 200mm gap size. Experimental Conditions were: Initial Wall Temperature = 200 -- 300 0 C, Back Pressure = 1 atm., Coolant Temperature = 71 -- 100 0 C, Extraction Water Velocity = 0 -- 2 cm/s, Gap Size = 50 mm. The effective water head history obtained in the experiment was compared with those predicted with Sudo's void fraction correlation. In the prediction, heat input to coolant was calculated from the response of measured wall temperature with heat condition analysis. The experimental results and analysis reveals that: (1) The effects of the gap size and initial wall temperature are evident, (2) The effect of extraction water velocity is negligible, and (3) The predicted history of effective water head is in good agreement with the experimental results except during the transient period in which the effective water head is descreasing. (author)

  16. Potential for boron dilution during small-break LOCAs in PWRs

    International Nuclear Information System (INIS)

    Nourbakhsh, H.P.; Cheng, Z.

    1995-01-01

    This paper documents the results of a scoping study of boron dilution and mixing phenomena during small break loss of coolant accidents (LOCAs) in pressurized water reactors (PWRs). Boron free condensate can accumulate in the cold leg loop seals when the reactor is operating in a reflux/boiler condenser mode. A problem may occur when the subsequent change in flow conditions such as loop seal clearing or re-establishment of natural circulation flow drive the diluted water in the loop seals into the reactor core without sufficient mixing with the highly borated water in the reactor downcomer and lower plenum. The resulting low boron concentration coolant entering the core may cause a power excursion leading to fuel failure. The mixing processes associated with a slow moving stream of diluted water through the loop seal to the core were examined in this report. A bounding evaluation of the range of boron concentration entering the core during a small break LOCA in a typical Westinghouse-designed, four-loop plant is also presented in this report

  17. Detection and control of potential core damage during a small-break LOCA

    International Nuclear Information System (INIS)

    Thomas, G.R.; Zebroski, E.L.

    1981-01-01

    A refreshing development in small-break LOCA analysis and testing is the recognition that this work can be of real value to a plant operator. Event-trees, or safety sequence diagrams, are being made increasingly realistic and are being used to develop and to test the abnormal transient operating guidelines (ATOGs) which provide a basis for operator response, training, and simulator work now under way. Perhaps the most monumental lesson of the TMI-2 accident is that the tradition of extreme worst-case analysis and data gathering can provide a directly negative contribution to safety if it is used as the basis for designing procedures and training of operator response. The event-trees for guiding operator response to abnormal conditions, ATOGs, must be based on physically realistic, best-estimate models. Possibly, the most dramatic risk reduction achieved will be attained through the use of realistic accident analysis, which leads to realistic operator guidelines and training, improved display of the critical information to the operator, and improved management structure. Given the dominant contribution of small-break LOCAs to the overall public risk envelope, realism should be the primary banner for future work in this field

  18. Water volume available for ECCS sump recirculation mode following a LOCA

    International Nuclear Information System (INIS)

    Riekert, T.; Rebohm, H.; Huber, J.; Brandes, F.

    2006-01-01

    In this paper we describe the reviews performed in Germany on the water level in the containment sump after a LOCA and the derived actions. Our view on the issue is from the perspective of the independent safety experts - i.e. TUV SUD Industrie Service (TUV SUD IS), TUV SUD Energietechnik GmbH Baden-Wuerttemberg (TUV SUD ET), TUV NORD EnSys Hannover and TUV NORD SysTec -, which reviewed the analyses of the utilities on behalf of the responsible supervising authorities. Between these expert organizations information were exchanged via the steering committee on nuclear technology of the association of the TUVs (VdTUV). In our paper we describe the analyses on the two safety issues relevant in the connection with the water level in the containment sump: the necessary minimum coverage of suction pipes to avoid inadmissible entrainment of air and the water retention inside the containment after a LOCA. Our description concentrates on PWRs because of the more complex conditions in comparison to BWRs. In conclusion it can be stated that due to the thorough evaluation of operating experience, optimization measures could be derived. In addition, the analyses served the purpose of know-how maintenance. (authors)

  19. Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

  20. Analysis of a LOCA crash into a reactor containment PWR-W with the GOTHIC code; Analisis de un accidente LOCA en contencion de un reactor PWR-W con el codigo GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Perianez Alvarez, V.

    2013-07-01

    The main objective of the work is the simulation of a severe accident, type LOCA with the GOTHIC-code, calculations of pressure and temperature in the containment levels, as well as the three-dimensional distribution of the inventory within the containment.

  1. Risk-Informed Margin Management (RIMM) Industry Applications IA1 - Integrated Cladding ECCS/LOCA Performance Analysis - Problem Statement

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yurko, Joseph P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swindlehurst, Gregg [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the LOCA/ECCS acceptance criteria to include the effects of higher burnup on cladding performance as well as to address some other issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in the summer of 2016. The impact of the final 50.46c rule on the industry will involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or reanalyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 5-10 years following the rule effective date. The need to use advanced cladding designs is expected. A loss of operational margin will result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin.

  2. Hydrogen radiolytic production in light and heavy water mixtures under conditions similar to LOCA (loss of coolant accidents)

    International Nuclear Information System (INIS)

    Garcia Rodenas, L.; Ali, S.P.; Liberman, S.J.

    1987-01-01

    H 2 , HD and D 2 radiolytic yield in heavy and light water mixtures has been determined to supply the necessary data which will allow to make a realistic estimation of the solution of such gas under LOCA conditions as a function of time. (Author)

  3. LOFT/L2-5, Loss of Fluid Test, 3. NRC L2 Large Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the third of the NRC L2 Series of nuclear large Break LOCA experiments, conducted on 16 June 1981. It simulated a 100% cold leg break with a maximum heat generation of 40 kW/m and rapid pump coast down

  4. BWR 2 % main recirculation line break LOCA tests RUNs 915 and 920 without HPCS in ROSA-III program

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Yonomoto, Taisuke; Koizumi, Yasuo; Tasaka, Kanji

    1987-03-01

    This report presents the experimental results of BWR LOCA integral tests, RUNs 915 and 920, which are performed in the ROSA-III program simulating 2 % main recirculation line break LOCA tests with and without pressure control system operation. The ROSA-III test facility simulates a BWR system with volume scale of 1/424 and has four half-length electrically heated fuel bundles, two active recirculation loops, four types of ECCS's, and steam and feedwater systems. The report presents (1) the experimental results of 2 % small break LOCA phanomena in the ROSA-III system and (2) the effects of the pressure control system on the LOCA phenomena. The pressure control system contributed to (A) prevent bulk flashing in the early blowdown phase, (B) early closure of MSIV by L2 level trip, (C) early actuation of ADS by L1 level trip. However, the core thermal responses of the two tests were similar because of the similar mass inventory in PV after the ADS actuation in both tests. (author)

  5. Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Youngmin; Kim, Young-In; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    KAERI (Korea Atomic Energy Research Institute) is developing a passive safety injection system (PSIS) to supply cold borated water into a reactor coolant system (RCS) without any operator actions or AC power under the occurrence of postulated design basis accidents. The PSIS consists of four independent trains, each of which is furnished with a gravity drained core makeup tank (CMT) and a safety injection tank (SIT). The CMT is designed to provide makeup and boration functions to the RCS during the early stage of a loss of coolant accident (LOCA) and a non-LOCA. In this paper, we investigate numerically the fluid mixing characteristics in the upper annular space of SMART, especially when single-phase natural circulation is formed between the CMT and RCS following a non-LOCA such as a main steam line break. In this paper, the fluid mixing characteristics in the upper annular space of SMART are investigated numerically when single-phase natural circulation is formed between the RCS and CMT during the early stage of a non-LOCA.

  6. Numerical Simulation of Fluid Mixing in Upper Annular Space of SMART during Early Stage of non-LOCA

    International Nuclear Information System (INIS)

    Bae, Youngmin; Kim, Young-In; Kim, Keung Koo

    2015-01-01

    KAERI (Korea Atomic Energy Research Institute) is developing a passive safety injection system (PSIS) to supply cold borated water into a reactor coolant system (RCS) without any operator actions or AC power under the occurrence of postulated design basis accidents. The PSIS consists of four independent trains, each of which is furnished with a gravity drained core makeup tank (CMT) and a safety injection tank (SIT). The CMT is designed to provide makeup and boration functions to the RCS during the early stage of a loss of coolant accident (LOCA) and a non-LOCA. In this paper, we investigate numerically the fluid mixing characteristics in the upper annular space of SMART, especially when single-phase natural circulation is formed between the CMT and RCS following a non-LOCA such as a main steam line break. In this paper, the fluid mixing characteristics in the upper annular space of SMART are investigated numerically when single-phase natural circulation is formed between the RCS and CMT during the early stage of a non-LOCA

  7. Korean Consortium's preliminary research for enhancing a probabilistic fracture mechanics code, PRO-LOCA

    International Nuclear Information System (INIS)

    Kim, Sun-Hye; Park, Jung-Soon; Lee, Jin-Ho; Yun, Eun-Sub; Kang, Sun-Ye; Shim, Do-Jun

    2015-01-01

    The Battelle developed a probabilistic fracture mechanics code called PRO-LOCA, which can be used as a tool for evaluating the pipe break frequency. It is being further developed through the international co-operative research program, PARTRIDGE. KINS, KHNP-CRI, and KEPCO-E&C are participating in the PARTIRDGE program by composing a Korean Consortium. The members of Korean Consortium performed benchmark analyses using the beta version of PRO-LOCA 4.0 to evaluate the effect of variables such as simulation methods, crack features, loading conditions, and inspection models on the failure probabilities. The benchmark analyses showed that the PRO-LOCA can provide a trend consistent with the expected crack growth and pipe failure behavior. Especially, the availability of the stress intensity factor and crack opening displacement for non-idealized through-wall cracks was proven from this study. This new solution for non-idealized through-wall cracks had been developed by the Korean Consortium and it was newly included in PRO-LOCA 4.0. However, further improvement is needed to address the problems such as the instability of adaptive sampling method and the unexpected trend of failure probabilities at the early stage of crack growth

  8. LOFT/LP-SB-3, Loss of Fluid Test, Cold Leg Break LOCA, No High Pressure injection System (HPIS)

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The sixth OECD LOFT experiment was conducted on 5 March 1984. It simulated a 1.8-in cold leg break LOCA with no HPIS available. This experiment was designed mainly for investigation of plant recovery effectiveness using secondary bleed and feed during core uncover and addressed accumulator injection at low pressure differentials. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  9. LOFT/L2-3, Loss of Fluid Test, 2. NRC L2 Large Break LOCA Experiment

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: This experiment was the second of the NRC L2 Series of nuclear large Break LOCA experiments, and was conducted on 12 May 1979. It simulated a 100% cold leg break with a maximum heat generation of 39 kW/m

  10. IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU

    International Nuclear Information System (INIS)

    Cunningham, Mitchel E.; Turnbull, J.A.

    2003-01-01

    Description - Objectives - Results: The U.S. Nuclear Regulatory Commission (NRC) conducted a series of thermal-hydraulic and cladding mechanical deformation tests in the National Research Universal (NRU) reactor at the Chalk River National Laboratory in Canada. The objective of these tests was to perform simulated loss-of-coolant-accident (LOCA) experiments using full-length light-water reactor fuel rods to study mechanical deformation, flow blockage, and coolability. Three phases of a LOCA (i.e., heat-up, reflood, and quench) were performed in situ using nuclear fissioning to simulate the low-level decay power during a LOCA after shutdown. All tests used PWR-type, non-irradiated fuel rods. Provided here is information on two materials tests, MT-6A and MT-4, which PNNL considers the better characterized for the purposes of setting up computer cases. The NRU reactor is a heterogeneous, thermal, tank-type research reactor. It has a power level of 135 MWth and is heavy-water moderated and cooled. The coolant has an inlet temperature of 310 K at a pressure of 0.65 MPa. The MT tests were conducted in a specially designed test train to supply the specified coolant conditions of flowing steam, stagnant steam, and then reflood. Typical instrumentation for the MT tests included fuel centerline thermocouples, cladding inner surface thermocouples, cladding outer surface thermocouples, rod internal gas pressure transducers or pressure switches, coolant channel steam probes, and self-powered neutron detectors. This instrumentation allowed for determining rupture times and cladding temperature. The test rods for the LOCA cases in the NRU reactor were irradiated in flowing steam prior to the transient, stagnant steam during the transient and prior to reflood, and then reflood conditions to complete the transient. Both cladding inner surface and outer surface temperatures were measured, in addition to coolant temperatures. However, only cladding inner surface temperatures were

  11. Large-break LOCA assessment for the highly advanced core design

    International Nuclear Information System (INIS)

    Doria, F.J.; Nath, V.I.; Hau, K.F.; Dam, R.F.; Vecchiarelli, J.

    1997-01-01

    Over the course of the years, a conceptual highly advanced core (HAC) reactor has been designed for Japan Electric Power Development Company Limited (EPDC). The HAC reactor, which is capable of generating 1326 MW of electrical power, consists of 640 CANDU-type fuel channels with each fuel channel containing twelve 61-element fuel bundles. As part of the conceptual design study, the performance of the HAC reactor during a large loss-of-coolant accident (LOCA) was assessed with the use of several computer codes. The SOPHT, CATHENA, ELOCA and ELESTRES computer codes were used to predict the thermalhydraulic behaviour of the circuit, thermalhydraulic behaviour of a single high-power channel, thermal-mechanical behaviour of the outer fuel elements contained in the high-powered channel and the steady-state fuel-element conditions respectively. The LOCAs that were analyzed include 100% reactor outlet header (ROH) break, and a survey of reactor inlet header (RIH) breaks ranging from 5% to 25%. The conceptual feasibility of the HAC design was evaluated against two criteria; namely, maximum sheath temperature less than 1200 deg C and AECL's 5% sheath straining criterion to assess failure by excessive straining. For the cases analyzed, the analysis predicted a maximum sheath temperature of 820 deg C and a maximum sheath strain of 1.5% (the maximum pressure-tube temperature was 515 deg C). Although the maximum element-burnup of the HAC design is extended beyond the CANDU 6 burnup, the maximum linear power of HAC (40 kW/m) is significantly lower than the maximum linear power of a CANDU 6 reactor (60 kW/m). The reduced element-power level in conjunction with internal design modification for the HAC design has resulted m significantly lower internal gas pressures under steady-state conditions, as compared with the CANDU 6 design. During a LOCA, the low linear powers and zero-void reactivity associated with the HAC design has increased the safety margin. In addition, the cases

  12. Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

    International Nuclear Information System (INIS)

    Sdouz, Gert

    2006-01-01

    The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

  13. Numerical Ballooning and Burst Prediction of Fuel Cladding During LOCA Transients in LWR

    International Nuclear Information System (INIS)

    Landau, E.; Weiss, Y.; Szanto, M.

    2014-01-01

    Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis, most former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. Several codes for reactor core and fuel rod analysis exist nowadays, such as FRAPTRAN1.4 or RELAP5-3D . These are tailor-made codes, designed to predict general core behavior and fuel performance, and while they are also used in predicting core components behavior during accident conditions, including those of cladding ballooning and failure with good accuracy, they contain several limitations on modeling the full transient cladding thermo mechanical phenomena. Limitations such as mechanical models being one dimensional or in axisymmetric geometries only, relying mostly on analytical models therefore having further restricting assumptions in return for accuracy, etc. These limitations disable the simulation of several important aspects, such as modeling 3D azimuthal behavior for example. The objective of the current work is to develop a comprehensive numerical model for predicting zircalloy cladding thermo mechanical behavior during a LOCA. The model will eventually predicts full cladding ballooning and burst behavior followed by fuel relocation, for fuel rods that can be subjected to 3D distributed flux. The model is fully three dimensional and is created using the commercial FEM numerical simulation software ABAQUS© applying

  14. In-pile experiments on fuel rod behavior during a LOCA

    International Nuclear Information System (INIS)

    Karb, E.; Pruessmann, M.; Sepold, L.

    1980-05-01

    This report describes the results of the Test Series F, Tests F 1 through F 5, in the in-pile experimental program with single rods in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The research is part of the Nuclear Safety Project's (PNS) fuel behavior program. The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA. The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR. The main parameter of the FR2-LOCA test program is the burnup. The F tests were perfomed from Oct. 25, 1977 to Nov. 22, 1977. They were the first tests in this program to use pre-irradiated fuel rods. The nominal burnup of the test rods was 20 000 MWd/t. During the transient test, the test rods were subjected to rod powers between 36 and 41 W/cm and were pressurized with He to hot internal pressures between 46 and 83 bar. The test rods during the heatup phase at pressures of 56, 53, 42, 72 and 60 bar, respectively. The burst temperatures were determined to be 890, 893, 932, 835 and 880 0 C for test F 1 through F 5. The maximum total circumferential elongations amount to 59, 38, 27, 34 and 41%, respectively. The F tests revealed a fragmentation of the fuel after the irradiation (prior to the tests) and a disintegration of the fuel pellet column after the transient tests due to cladding ballooning. The post-test results indicated a significant reduction of the pellet stack length for all five test rods. The burst data of the F tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. (orig./HP) [de

  15. Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

    International Nuclear Information System (INIS)

    Fischer, S.R.; Farman, R.F.; Birdsell, S.A.

    1992-01-01

    This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model

  16. QUENCH-LOCA program at KIT and results of the QUENCH-L0 bundle test

    International Nuclear Information System (INIS)

    Stuckert, J.; Grosse, M.; Roessger, C.; Steinbrueck, M.; Walter, M.

    2012-01-01

    The current LOCA criteria and their safety goals are applied worldwide with minor modifications since the USNRC release in 1973. The criteria are given as limits on peak cladding temperature (T PCT ≤ 1200 C) and on oxidation level ECR (equivalent cladding reacted) calculated as a percentage of cladding oxidized (ECR ≤ 17% calculated using Baker-Just oxidation correlation). These two rules constitute the criterion of cladding embrittlement due to oxygen uptake. The results elaborated worldwide in the 1980s and 1990s on Zircaloy-4 (Zry-4) cladding tubes behavior (oxidation, deformation and bundle coolability) under LOCA conditions constitute a detailed data base and an important input for the safety assessment of LWRs. In-pile test data (with burn-up up to 35 MWd/kgU) were consistent with the out-of-pile data and did not indicate an influence of the nuclear environment on cladding deformation. At high burn-up, fuel rods fabricated from conventional Zry-4 often exhibit significant oxidation, hydriding, and oxide spallation. Thus, many fuel vendors have proposed the use of recently developed cladding alloys, such as M5 registered , ZIRLO trademark and other. Therefore, it is important to verify the safety margins for high burn-up fuel and fuel claddings with new alloys. Due to long cladding hydriding period for the high fuel burn-up, post-quench ductility is strongly influenced not only by oxidation but also hydrogen uptake. The 17% ECR limit is inadequate to ensure post-quench ductility at hydrogen concentrations higher than ∼500 wppm. Due to so called secondary hydriding (during oxidation of inner cladding surface after burst), which was firstly observed in JAEA, the hydrogen content can reach 4000 wppm in Zircaloy cladding regions around burst. To investigate the influence of these phenomena on the applicability of the embrittlement criteria for the German nuclear reactors it was decided to perform the QUENCH-LOCA bundle test series at the Karlsruhe Institute

  17. QUENCH-LOCA program at KIT and results of the QUENCH-L0 bundle test

    Energy Technology Data Exchange (ETDEWEB)

    Stuckert, J.; Grosse, M.; Roessger, C.; Steinbrueck, M.; Walter, M. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2012-11-01

    The current LOCA criteria and their safety goals are applied worldwide with minor modifications since the USNRC release in 1973. The criteria are given as limits on peak cladding temperature (T{sub PCT} {<=} 1200 C) and on oxidation level ECR (equivalent cladding reacted) calculated as a percentage of cladding oxidized (ECR {<=} 17% calculated using Baker-Just oxidation correlation). These two rules constitute the criterion of cladding embrittlement due to oxygen uptake. The results elaborated worldwide in the 1980s and 1990s on Zircaloy-4 (Zry-4) cladding tubes behavior (oxidation, deformation and bundle coolability) under LOCA conditions constitute a detailed data base and an important input for the safety assessment of LWRs. In-pile test data (with burn-up up to 35 MWd/kgU) were consistent with the out-of-pile data and did not indicate an influence of the nuclear environment on cladding deformation. At high burn-up, fuel rods fabricated from conventional Zry-4 often exhibit significant oxidation, hydriding, and oxide spallation. Thus, many fuel vendors have proposed the use of recently developed cladding alloys, such as M5 {sup registered}, ZIRLO trademark and other. Therefore, it is important to verify the safety margins for high burn-up fuel and fuel claddings with new alloys. Due to long cladding hydriding period for the high fuel burn-up, post-quench ductility is strongly influenced not only by oxidation but also hydrogen uptake. The 17% ECR limit is inadequate to ensure post-quench ductility at hydrogen concentrations higher than {approx}500 wppm. Due to so called secondary hydriding (during oxidation of inner cladding surface after burst), which was firstly observed in JAEA, the hydrogen content can reach 4000 wppm in Zircaloy cladding regions around burst. To investigate the influence of these phenomena on the applicability of the embrittlement criteria for the German nuclear reactors it was decided to perform the QUENCH-LOCA bundle test series at the

  18. Development of loca calculation capability with relap5-3D in accordance with the evaluation model methodology

    International Nuclear Information System (INIS)

    Liang, T.K.S.; Huan-Jen, Hung; Chin-Jang, Chang; Lance, Wang

    2001-01-01

    In light water reactors, particularly the pressurized water reactor (PWR), the severity of a LOCA (loss of coolant accident) will limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the PCT (peak cladding temperature) evaluation during LOCA, it generally takes more resources to develop. Instead, implementation of evaluation models required by the Appendix K of 10 CFR 50 upon an advanced thermal-hydraulic platform can also enlarge significant margin between the highest calculated PCT and the safety limit of 2200 F. The compliance of the current RELAP5-3D code with Appendix K of 10 CFR50 has been evaluated, and it was found that there are ten areas where code assessment and/or further modifications were required to satisfy the requirements set forth in the Appendix K of 10 CFR 50. The associated models for LOCA consequent phenomenon analysis should follow the major concern of regulation and be expected to give more conservative results than those by the best-estimate methodology. They were required to predict the decay power level, the blowdown hydraulics, the blowdown heat transfer, the flooding rate, and the flooding heat transfer. All of the ten areas included in above classified simulations have been further evaluated and the RELAP5-3D has been successfully modified to fulfill the associated requirements. In addition, to verify and assess the development of the Appendix K version of RELAP5-3D, nine separate-effect experiments were adopted. Through the assessments against separate-effect experiments, the success of the code modification in accordance with the Appendix K of 10 CFR 50 was demonstrated. We will apply another six sets of integral-effect experiments in the next step to assure the integral conservatism of the Appendix K version of RELAP5-3D on LOCA licensing evaluation. (authors)

  19. Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

  20. Space-time neutronic analysis of postulated LOCA's in CANDU reactors

    International Nuclear Information System (INIS)

    Luxat, J.C.; Frescura, G.M.

    1978-01-01

    Space-time neutronic behaviour of CANDU reactors is of importance in the analysis and design of reactor safety systems. A methodology has been developed for simulating CANDU space-time neutronics with application to the analysis of postulated LOCA'S. The approach involves the efficient use of a set of computer codes which provide a capability to perform simulations ranging from detailed, accurate 3-dimensional space-time to low-cost survey calculations using point kinetics with some ''effective'' spatial content. A new, space-time kinetics code based upon a modal expansion approach is described. This code provides an inexpensive and relatively accurate scoping tool for detailed 3-dimensional space-time simulations. (author)

  1. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

  2. Ethylene propylene cable degradation during LOCA research tests: tensile properties at the completion of accelerated aging

    International Nuclear Information System (INIS)

    Bustard, L.D.

    1982-05-01

    Six ethylene-propylene rubber (EPR) insulation materials were aged at elevated temperature and radiation stress exposures common in cable LOCA qualification tests. Material samples were subjected to various simultaneous and sequential aging simulations in preparation for accident environmental exposures. Tensile properties subsequent to the aging exposure sequences are reported. The tensile properties of some, but not all, specimens were sensitive to the order of radiation and elevated temperature stress exposure. Other specimens showed more severe degradation when simultaneously exposed to radiation and elevated temperature as opposed to the sequential exposure to the same stresses. Results illustrate the difficulty in defining a single test procedure for nuclear safety-related qualification of EPR elastomers. A common worst-case sequential aging sequence could not be identified

  3. The fuel and channel thermal/mechanical behaviour code FACTAR 2.0 (LOCA)

    International Nuclear Information System (INIS)

    Westbye, C.J.; Mackinnon, J.C.; Gu, B.W.

    1996-01-01

    The computer code FACTAR 2.0 (LOCA) models the thermal and mechanical response of components within a single CANDU fuel channel under loss-of-coolant accident conditions. This code version is the successor to the FACTAR 1.x code series, and features many modelling enhancements over its predecessor. In particular, the thermal hydraulic treatment has been extended to model reverse and bi-directional coolant flow, and the axial variation in coolant flow rate. Thermal radiation is calculated by a detailed surface-to-surface model, and the ability to represent a greater range of geometries (including experimental configurations employed in code validation) has been implemented. Details of these new code treatments are described in this paper. (author)

  4. Large break LOCA analysis for retrofitted ECCS at MAPS using modified computer code ATMIKA

    International Nuclear Information System (INIS)

    Singhal, Mukesh; Khan, T.A.; Yadav, S.K.; Pramod, P.; Rammohan, H.P.; Bajaj, S.S.

    2002-01-01

    Full text: Computer code ATMIKA which has been used for thermal hydraulic analysis is based on unequal velocity equal temperature (UVET) model. Thermal hydraulic transient was predicted using three conservation equations and drift flux model. The modified drift flux model is now able to predict counter current flow and the relative velocity in vertical channel more accurately. Apart from this, stratification model is also introduced to predict the fuel behaviour under stratified condition. Many more improvements were carried out with respect to solution of conservation equation, heat transfer package and frictional pressure drop model. All these modifications have been well validated with published data on RD-12/RD-14 experiments. This paper describes the code modifications and also deals with the application of the code for the large break LOCA analysis for retrofitted emergency core cooling system (ECCS) being implemented at Madras Atomic Power Station (MAPS). This paper also brings out the effect of accumulator on stratification and fuel behaviour

  5. Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

    International Nuclear Information System (INIS)

    Jiang, J.T.; Luxat, J.C.

    2008-01-01

    Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

  6. Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

    2008-07-01

    Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

  7. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    International Nuclear Information System (INIS)

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-01-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  8. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

  9. Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs

    International Nuclear Information System (INIS)

    Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

    2013-01-01

    SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

  10. Reactor elements properties response during a postulated loss-of-coolant accident (LOCA)

    International Nuclear Information System (INIS)

    Ahmed, E.E.; Rahman, F.A.

    1985-01-01

    Four computer algorithms have been introduced to solve for the reactor different materials response subjected to LOCA conditions, they were developed with the intent of producing a simple, accurate and efficient prediction schemes. A general overview of the solution procedures design and working of each of four algorithms are presented, followed by short description of the nature of solution and calculated results. These algorithms are: 1. ZIRCP to give the cladding material properties response under normal and transient conditions. 2. FCGAPP to give the fuel- cladding gas-gap conductivity. 3. NFUEIP to solve the temperature dependent of nuclear fuel properties during normal and transient conditions. 4. TSDATP has been developed to solve for the thermodynamic and transport properties of water and steam over a large range of temperature and pressure. 14 fig

  11. TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

  12. Reentrainment of droplet from grid spacer in mist flow portion of LOCA reflood of PWR

    International Nuclear Information System (INIS)

    Lee, S.L.; Cho, S.K.; Sheen, H.J.

    1983-01-01

    An investigation is made on the influence of a quenched grid spacer on the greatly enhanced heat transfer from heated fuel rods during the mist flow phase of emergency reflood of loss of coolant accident (LOCA) of a pressurized water reactor (PWR). The situation for the case of a dry grid spacer before its quenching has not been covered in this study. The experimental technique used is a relatively simple optical scheme which combines the reference-mode laser-Doppler anemometry making use of the scattering of a light beam from a droplet. The results reveal that the large droplets in the mist flow, which are intercepted by the grid spacer, are responsible for the creation of a large number of smaller droplets. These small droplets, due to their large surface area to mass ratios, can serve as superb evaporative cooling agents to heat transfer downstream of the grid spacer

  13. Estimation of the core-wide fuel rod damage during a LWR LOCA

    International Nuclear Information System (INIS)

    Mattila, L.; Sairanen, R.; Stengaard, J.-O.

    1975-01-01

    The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

  14. Thermal stresses at nozzles of nuclear steel containments under LOCA-conditions

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, G.; Bergmann, A.N.

    1986-01-01

    During a loss of coolant accident (LOCA) of a PWR-nuclear power plant, a considerable heating of the containment atmosphere is expected to occur. Transient thermal stresses will appear at the containment as a consequence of a non-uniform rise of its temperature. Applying computer codes based on the finite element method, dimensionless general thermal stresses at nozzles of spherical steel containment have been calculated, varying the principal geometrical parameters and the Biot number for the containment internal surface. Atmosphere temperature and Biot number are assumed constant after the accident. Several plots of the maximum principal stresses are provided, which constitute general results applicable to stress analysis of any particular containment of this kind. (orig.)

  15. Hydrogen and steam distribution following a small-break LOCA in large dry containment

    Institute of Scientific and Technical Information of China (English)

    DENG Jian; CAO Xuewu

    2007-01-01

    The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations.Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident (LOCA) without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump (RCP)compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.

  16. Verification of human actions in SBO sequences with LOCA stamps in Westinghouse PWRs

    International Nuclear Information System (INIS)

    Queral, C.; Mena Rosell, L.; Jimenez Varas, G.

    2013-01-01

    The Fukushima accident has shown the need for tools and methodologies able to analyze human activities and / or capabilities of portable systems that has given the Spanish plants as a result of the stress tests . In this work we have applied the methodology of integrated safety analysis developed by the CSN , to SBO sequences with LOCA stamp. The aim is to show a methodology for testing the performances of the Emergency Operating Procedures and Guides Severe Accident Management. The simulations were performed with the tool SCAIS coupled to MAAP . The results show that there are human activities that may be beneficial in certain sequences but harmful in others. This type of problem is already known and referred to in the GGAS . However, FSR shows a practical way to check human actions cannot be obtained with other methods.

  17. Fast instrumentation for loss of coolant accident (LOCA) experimental studies pertaining to nuclear reactors

    International Nuclear Information System (INIS)

    Venkat Raj, V.; Sreenivas Rao, G.; Belokar, D.G.; Dolas, P.K.

    1989-01-01

    The loss of coolant accident (LOCA) which involves a breach in the pressure boundary of the primary coolant system (PCS) is one of the postulated accident conditions against which the safety of the reactor system is to be ensured. Mathematical models have been developed to analyse this kind of transients. However, because of the extremely complicated nature of the phenomena involved, it is necessary to validate the analytical models with appropriate experimental data. Many parameters are to be measured during the experiments, out of which temperature, pressure, void fraction and two-phase mass flow rate are the most important parameters. Since the phenomenon is very fast, special fast response instruments are required. This paper deals with the considerations that govern the selection of appropriate instruments and the development of suitable instruments for transient two-phase flow and void fraction measurements. The requirements of the associated fast data acquisition system are also discussed. (author). 4 figs

  18. LOCA, LOFA and LOVA analyses pertaining to NET/ITER safety design guidance

    International Nuclear Information System (INIS)

    Ebert, E.; Raeder, J.

    1991-01-01

    The analyses presented pertain to loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA). These types of accidents may jeopardise components and plasma vessel integrity and cause radioactivity mobilisation. The analyses reviewed have been performed under the assumption that the plasma facing components are protected by a carbon based armour. Accidental temperatures and pressure transients are quantified, the possibility of reaction products combustion is investigated and worst case accidental public doses are assessed. On this basis, design recommendations are given and design features such as low plasma facing components armour temperatures (on almost the entire surface) and inert gas adjacent to the vacuum vessel have been implemented. (orig.)

  19. OECD-LOFT large break LOCA experiments: phenomenology and computer code analyses

    International Nuclear Information System (INIS)

    Brittain, I.; Aksan, S.N.

    1990-08-01

    Large break LOCA data from LOFT are a very important part of the world database. This paper describes the two double-ended cold leg break tests LP-02-6 and LP-LB-1 carried out within the OECD-LOFT Programme. Tests in LOFT were the first to show the importance of both bottom-up and top-down quenching during blowdown in removing stored energy from the fuel. These phenomena are discussed in detail, together with the related topics of the thermal performance of nuclear fuel and its simulation by electric fuel rod simulators, and the accuracy of cladding external thermocouples. The LOFT data are particularly important in the validation of integral thermal-hydraulics codes such as TRAC and RELAP5. Several OECD partner countries contributed analyses of the large break tests. Results of these analyses are summarised and some conclusions drawn. 32 figs., 3 tabs., 45 refs

  20. Small break LOCA RELAP5/MOD3 uncertainty quantification: Bias and uncertainty evaluation for important phenomena

    International Nuclear Information System (INIS)

    Ortiz, M.G.; Ghan, L.S.; Vogl, J.

    1991-01-01

    The Nuclear Regulatory Commission (NRC) revised the Emergency Core Cooling System (ECCS) licensing rule to allow the use of Best Estimate (BE) computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability and Uncertainty (CSAU) to evaluate BE code uncertainties. The CSAU methodology was demonstrated with a specific application to a pressurized water reactor (PWR), experiencing a postulated large break loss-of-coolant accident (LBLOCA). The current work is part of an effort to adapt and demonstrate the CSAU methodology to a small break (SB) LOCA in a PWR of B and W design using RELAP5/MOD3 as the simulation tool. The subject of this paper is the Assessment and Ranging of Parameters (Element 2 of the CSAU methodology), which determines the contribution to uncertainty of specific models in the code

  1. Investigation of bubble-condenser operation under large break LOCA conditions

    International Nuclear Information System (INIS)

    Blinkov, V.; Melikhov, O.; Melikhov, V.; Davydov, M.; Sokolin, A.; Hoffmann, D.; Simon, U.; Bajsz, J.

    2000-01-01

    In the framework of the PHARE/TACIS project, the experimental test facility for bubble condenser experimental qualification was built at Electrogorsk Research and Engineering Centre. The test facility contains high pressure system, compartments upstream of the bubble condenser and a section of the bubble condenser system. The scaling of the test facility is 1:100. The high pressure system consists of five vessels to appropriately model the leak functions (mass flow rate and enthalpy) during the loss of coolant accidents postulated in the design of VVER-440/V213. Design basis accident (LB LOCA) was experimentally and analytically considered. Results of pre-test analysis with ATHLET and DRASYS codes for determination of necessary test parameters and post-test analysis of three tests are presented. (author)

  2. Prediction of the fuel failure following a large LOCA using modified gap heat transfer model

    International Nuclear Information System (INIS)

    Lee, K.M.; Lee, N.H.; Huh, J.Y.; Seo, S.K.; Choi, J.H.

    1995-01-01

    The modified Ross and Stoute gap heat transfer model in the ELOCA.Mk5 code for CANDU safety analysis is based on a simplified thermal deformation model. A review on a series of recent experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this study, more realistic offset crap conductance model is implemented in the code to estimate the fuel failure thresholds usincr the transient conditions of a 100% Reactor Outlet Header (ROH) break LOCA. Based on the offset gap conductance model, the total release of I-131 from the failed fuel elements in the core is reduced from 3876 TBq to 3283 TBq to increase margin for dose limit. (author)

  3. Analysis of a large-break LOCA at lower operational modes

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, H.Y.; Jun, H.Y.; Lee, K. [Korea Electric Power Corporation, Taejon (Korea)

    2000-10-01

    To improve Technical Specifications and Emergency Operating Guidelines (EOGs) applicable at lower operational modes it is required to perform the safety analysis reflecting the operational characteristics in those modes. Because the component availability and system configurations at lower modes are different from those of power mode, the plant safety at lower modes should be confirmed through independent analyses. In the present study, a large-break loss-of-coolant accident is analyzed to evaluate the containment pressure and temperature control function for the preparation of EOGs applicable at lower modes. To reach the required shutdown condition, the plant cool-down is controlled by the secondary steam flow and auxiliary feedwater. The mass and energy releases from primary system are obtained from RELAP5/MOD3.1 calculation and the containment pressure and temperature are evaluated with CONTEMPT-LT code. The reference plant is Korean Next Generation Reactor having 4,000 MW thermal power. Two cases of cold leg LOCA initiated at Mode 3 with and without SIT operation are calculated. At the given plant conditions, all safety injection pumps are still available. The calculation at the condition of maximum mass and energy release shows that the containment pressure and temperature can be controlled within acceptable criteria, which means the operations of 2 or 4 fan coolers are the possible success paths to achieve the containment P/T control safety function. The peak cladding temperature with minimum safety injection flow does not show remarkable excursion, which implies the lower mode LOCA at Mode 3 can be bounded by the results obtained at full power from the viewpoint of ECCS performance. (author)

  4. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    Energy Technology Data Exchange (ETDEWEB)

    Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  5. Concept of the LORELEI Test Device for LOCA Experiment in the JHR Reactor

    International Nuclear Information System (INIS)

    Moran, N.; Ferry, L.; Azulay, A.; Mileguir, O.; Weiss, Y.; Szanto, M.

    2014-01-01

    Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis. Former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. In JHR material testing reactor, which is currently under construction, one significant experimental device is the LORELEI testing device. The objective is to examine the LOCA sequence influence on: thermo-mechanical behavior of the fuel clad, possible fuel relocation, corrosion at high temperature, oxidation, hydriding and resulted clad embrittment. The device is a single rod closed loop system placed on a displacement device inside a defined channel in the reflector. Several operational constrains on the device, as required by the reactor operational philosophy resulted quite a few challenges in the design. Constrains as: pre experimental re-irradiation phase under thermo-syphonic flow, application of active insulation to simulate the surrounding fuel, application of tensile force during refolding simulation, controlling the experiment with non-direct temperature measurement, etc. requires sophisticated solutions. The main objective of the conceptual design was to remove the uncertainties of those challenging requirements. The current presentation describes the approach applied defining the concept of the device, using sophisticated design combined with computational and experimental tools

  6. ASTEC and MELCOR comparison for a VVER-1000 60 mm small break LOCA

    International Nuclear Information System (INIS)

    Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

    2005-01-01

    In this paper a comparison between severe accident calculations performed for a WWER 1000 with the ASTEC1.1v0 and MELCOR 1.8.5 computer codes for a small break LOCA (ID 60 mm) without intervention of hydro accumulators is presented. This investigation has been performed in the framework of the SARNET project under the EURATOM 6th framework program. Once the accident sequence scenario is specified, both codes (MELCORE and ASTEC) are able to determine the core and containment damaged states, to estimate the release of radionuclides from the fuel as well as from the primary circuit and containment. Theses results are used to estimate the maximum period of the time during which the personnel could still take particular decisions in order to mitigate such an accident. The aim of the performed analysis is to estimate the discrepancy between ASTEC and MELCORE 1.8.5 calculations. Such discrepancies will be studied, if the case, proposal for ASTEC improvements will be made. Also the ASTEC capability to simulate specific reactor accident scenarios and/or particular safety systems will be tested. The final target is to propose severe accident management procedure for WWER 1000 reactors. In conclusions, the analysis for a small break LOCA (ID 60 mm without hydroelectricities) has shown some discrepancies between ASTEC and MELCORE especially during the degradation of the core. Further analyses are planed in which the MELCORE temperature 'set point' for core degradation (2520 K) will be progressively increased to approach the ASTEC one (which has been estimated to be about 3200 K). The comparison of the new results will allow a better evaluation of the in-vessel models implemented in ASTEC

  7. Estimation of Leak Flow Rate during Post-LOCA Using Cascaded Fuzzy Neural Networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    In this study, important parameters such as the break position, size, and leak flow rate of loss of coolant accidents (LOCAs), provide operators with essential information for recovering the cooling capability of the nuclear reactor core, for preventing the reactor core from melting down, and for managing severe accidents effectively. Leak flow rate should consist of break size, differential pressure, temperature, and so on (where differential pressure means difference between internal and external reactor vessel pressure). The leak flow rate is strongly dependent on the break size and the differential pressure, but the break size is not measured and the integrity of pressure sensors is not assured in severe circumstances. In this paper, a cascaded fuzzy neural network (CFNN) model is appropriately proposed to estimate the leak flow rate out of break, which has a direct impact on the important times (time approaching the core exit temperature that exceeds 1200 .deg. F, core uncover time, reactor vessel failure time, etc.). The CFNN is a data-based model, it requires data to develop and verify itself. Because few actual severe accident data exist, it is essential to obtain the data required in the proposed model using numerical simulations. In this study, a CFNN model was developed to predict the leak flow rate before proceeding to severe LOCAs. The simulations showed that the developed CFNN model accurately predicted the leak flow rate with less error than 0.5%. The CFNN model is much better than FNN model under the same conditions, such as the same fuzzy rules. At the result of comparison, the RMS errors of the CFNN model were reduced by approximately 82 ~ 97% of those of the FNN model.

  8. Revisiting large break LOCA with the CATHARE-3 three-field model

    International Nuclear Information System (INIS)

    Valette, Michel; Pouvreau, Jérôme; Bestion, Dominique; Emonot, Philippe

    2011-01-01

    Highlights: ► CATHARE 3 enables a three-field analysis of a LB LOCA. ► Reflooding experiments in isolated rod bundles are satisfactory predicted. ► A BETHSY integral test simulation supports the CATHARE 3 3-field assessment. - Abstract: Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in a rod bundle is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit: core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic six-equation model is used in the other parts of the loop. An analysis of these first results is presented and future work is defined for improving the droplet behavior simulation in both the upper plenum and the hot legs.

  9. LOFT/LP-LB-1, Loss of Fluid Test, Large-Break LOCA Experiment

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, Thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCE is expected to closely model a LPWR LOCA. 2 - Description of test: Experiment LP-LB-1 was conducted on 3 February 1984 in the Loss-of-Fluid Test (LOFT) facility at the Idaho National Engineering Laboratory under the auspices of the Organization for Economic Cooperation and Development. The primary objectives of Experiment LP-LB-1 were to determine system transient characteristics and to assess code predictive capabilities for design basis large-break loss-of-coolant accidents in pressurized water reactors (PWRs). This experiment simulated a double-ended offset shear of one inlet pipe in a four-loop PWR and was initiated from conditions representative of licensing limits in a PWR. Other boundary conditions for the simulation were loss of offsite power, rapid primary coolant pump coast down, and United Kingdom minimum safeguard emergency core coolant injection rates. The nuclear fuel rods were not pressurized. The transient was initiated by opening the quick-opening blowdown valves in the broken loop hot and cold legs. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  10. Hydrodynamics of AHWR gravity driven water pool under simulated LOCA conditions

    International Nuclear Information System (INIS)

    Thangamani, I.; Verma, Vishnu; Ali, Seik Mansoor

    2015-01-01

    The Advanced Heavy Water Reactor (AHWR) employs a double containment concept with a large inventory of water within the Gravity Driven Water Pool (GDWP) located at a high elevation within the primary containment building. GDWP performs several important safety functions in a passive manner, and hence it is essential to understand the hydrodynamics that this pool will be subjected to in case of an accident such as LOCA. In this paper, a detailed thermal hydraulic analysis for AHWR containment transients is presented for postulated LOCA scenarios involving RIH break sizes ranging from 2% to 50%. The analysis is carried out using in-house containment thermal hydraulics code 'CONTRAN'. The blowdown mass and energy discharge data for each break size, along with the geometrical details of the AHWR containment forms the main input for the analysis. Apart from obtaining the pressure and temperature transients within the containment building, the focus of this work is on simulating the hydrodynamic phenomena of vent clearing and pool swell occurring in the GDWP. The variation of several key parameters such as primary containment V1 and V2 volume pressure, temperature and V1-V2 differential pressure with time, BOP rupture time, vent clearing velocity, effect of pool swell on the V2 air-space pressure, GDWP water level etc. are discussed in detail and important findings are highlighted. Further, the effect of neglecting the pool swell phenomenon on the containment transients is also clearly brought out by a comparative study. The numerical studies presented in this paper give insight into containment transients that would be useful to both the system designer as well as the regulator. (author)

  11. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    International Nuclear Information System (INIS)

    Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

    2017-01-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  12. Comparison report for CSNI International Standard Problem 12 (ROSA-III Run 912)

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Iriko, Masanori; Yonomoto, Taisuke; Shiba, Masayoshi

    1982-09-01

    ROSA-III Run 912 was identified as International Standard Problem 12 by the Committee on the Safety of Nuclear Installations. Run 912 simulated a 5% split break LOCA condition in a BWR at the pump suction in the recirculation line with the HPCS failure. Comparisons between the test data and the calculations by eight international participants were made and discussed. (author)

  13. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  14. Synopsis of the results of ISP-39 on FARO test L-14

    Energy Technology Data Exchange (ETDEWEB)

    Addabbo, C.; Annunziato, A.; Magallon, D. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    This paper provides a synopsis of the salient results from the ISP(International Standard Problem)-39 exercise promoted by OECD-CSNI in the frame of the NEA activities aimed at fostering international cooperation in reactor safety research and development. ISP-39 has been conceived to benchmark the predictive capabilities of computer codes used in the evaluation of FCI and quenching phenomenologies of relevance in water cooled reactors severe accidents safety analysis. The ISP-39 reference case is FARO test L-14, a non-energetic FCI test performed in the FARO experimental installation under realistic melt composition and prototypical test conditions. (author)

  15. Impact of 2D/3D-project on LOCA-licensing analysis and reactor safety of PWRs

    International Nuclear Information System (INIS)

    Winkler, F.; Krebs, W.D.

    1989-01-01

    In the past LOCA-licensing analysis has included large conservatisms to compensate for the lack of detailed two phase flow and full scale experimental data. The 2D/3D-project was established to improve the data base in order to minimize the conservatisms required. The significant results and findings of the full scale Upper Plenum Test Facility (UPTF) and from the electrically heated Slab Core Test Facility (SCTF) were particularly useful for understanding the multidimensional phenomena in the primary system and in the core of a PWR. UPTF results were used to verify the TRAC-PF1 analysis of a PWR with combined ECC-Injection during the reflood phase of a large break-LOCA. Comparison of these results with results from classic licensing calculations quantifies the large safety margin in earlier licensing procedures and in reactor systems. (orig.)

  16. On-line pressurizer surveillance system design to prevent small break LOCA through PORV using micro-computer

    International Nuclear Information System (INIS)

    Lee, Jong-Ho; Chang, Soon-Heung

    1986-01-01

    Small break LOCA caused by a stuck-open PORV is one of the important contributors to nuclear power plant risk. This paper deals with the design of a pressurizer surveillance system using micro-computer to prevent the malfunction of system and has assessed the effect of this improvement through Probabilistic Risk Assessment (PRA) method. Micro-computer diagnoses the malfunction of system by a process checking method and performs automatically backup action related to each malfunction. Owing to this improvement, we can correctly diagnose ''Spurious Opening'', ''Fail to Reclose'' and ''Small break LOCA'' which are difficult for operator to diagnose quickly and correctly and reduce the probability of a human error by an automatic backup action. (author)

  17. Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

  18. Results of small break LOCA analysis for Kuosheng nuclear power plant using the RELAP5YA computer code

    International Nuclear Information System (INIS)

    Wang, L.C.; Jeng, S.C.; Chung, N.M.

    2004-01-01

    One lesson learned from the Three Mile Island (TMI) accident was the analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or nuclear fuel suppliers for small break Loss Of Coolant Accident (LOCA) analysis for compliance with appendix K to 10CFR50 should be revised, documented and submitted for USNRC approval and the plant-specific calculations using NRC-approved models for small-break LOCA to show compliance with 10CFR50.46 should be submitted for NRC approval. A study by Taiwan Power Company (TPC) under the guidance of Yankee Atomic Electric Company (YAEC) has been undertaken to perform this analysis for Kuosheng nuclear power plant. This paper presents the results of the analysis that are useful in satisfying the same requirements of the Republic Of China Atomic Energy Commission (ROCAEC). (author)

  19. Sensitivity analysis of local uncertainties in large break loss-of-coolant accident (LB-LOCA) thermo-mechanical simulations

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Ikonen, Timo

    2016-08-15

    Highlights: • A sensitivity analysis using the data from EPR LB-LOCA simulations is done. • A procedure to analyze such complex data is outlined. • Both visual and quantitative methods are used. • Input factors related to core design are identified as most significant. - Abstract: In this paper, a sensitivity analysis for the data originating from a large break loss-of-coolant accident (LB-LOCA) analysis of an EPR-type nuclear power plant is presented. In the preceding LOCA analysis, the number of failing fuel rods in the accident was established (Arkoma et al., 2015). However, the underlying causes for rod failures were not addressed. It is essential to bring out which input parameters and boundary conditions have significance to the outcome of the analysis, i.e. the ballooning and burst of the rods. Due to complexity of the existing data, the first part of the analysis consists of defining the relevant input parameters for the sensitivity analysis. Then, selected sensitivity measures are calculated between the chosen input and output parameters. The ultimate goal is to develop a systematic procedure for the sensitivity analysis of statistical LOCA simulation that takes into account the various sources of uncertainties in the calculation chain. In the current analysis, the most relevant parameters with respect to the cladding integrity are the decay heat power during the transient, the thermal hydraulic conditions in the rod’s location in reactor, and the steady-state irradiation history of the rod. Meanwhile, the tolerances in fuel manufacturing parameters were found to have negligible effect on cladding deformation.

  20. Effect of spray on performance of the hydrogen mitigation system during LB-LOCA for CPR1000 NPP

    International Nuclear Information System (INIS)

    Huang, X.G.; Yang, Y.H.; Cheng, X.; Al-Hawshabi, N.H.A.; Casey, S.P.

    2011-01-01

    Highlights: → This paper presents the spray effect on HMS during LB-LOCA by using GASFLOW. → The positive and negative effects of spray are summarized. → And the combination of DIS and PAR system is suggested as reasonable countermeasures. → This research is an important work aimed at the study of spray and hydrogen mitigation. → The contents of this paper should become a required part of the safety analysis of Chinese NPPs. - Abstract: During the course of the hypothetical large break loss-of-coolant accident (LB-LOCA) in a nuclear power plant (NPP), hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel (RPV). It is then ejected from the break into the containment along with a large amount of steam. Management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen mitigation system (HMS) and spray system in CPR1000 NPP. The computational fluid dynamics (CFD) code GASFLOW is utilized in this study to analyze the spray effect on the performance of HMS during LB-LOCA. Results show that as a kind of HMS, deliberate igniter system (DIS) could initiate hydrogen combustion immediately after the flammability limit of the gas mixture has been reached. However, it will increase the temperature and pressure drastically. Operating the DIS under spray condition could result in hydrogen combustion being suppressed by suspended droplets inside the containment. Furthermore, the droplets could also mitigate local the temperature rise. Operation of a PAR system, another kind of HMS, consumes hydrogen steadily with a lower recombination rate which is not affected noticeably by the spray system. Numerical results indicate that the dual concept, namely the integrated application of DIS and PAR systems, is a constructive improvement for hydrogen safety under spray condition during LB-LOCA.

  1. Calculation of BETHSY 0.5% small break LOCA with RELAP5-ISP 27 international activity of code assessment

    International Nuclear Information System (INIS)

    Chen Yuzhen

    1992-01-01

    BETHSY facility constructed in France is a 1/100 volumetrically-scaled full-pressure model of a PWR with 3 loops. ISP-27 is an international activity sponsored by OECD Nuclear Energy Agency. The experiment is a transient of 0.5% coldleg break LOCA with failure of HPIS. The calculations were performed with RELAP5/MOD2/36.05 at CYBER-170/825, which can present a good calculation, provided that the break flow is well modelled

  2. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    International Nuclear Information System (INIS)

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken

  3. Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

    International Nuclear Information System (INIS)

    Griggs, D.P.

    1993-03-01

    The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

  4. CONTEMPT-4MOD3, LWR Containment Long-Term Pressure Distribution and Temperature Distribution in LOCA

    International Nuclear Information System (INIS)

    Lin, C.C.; Economos, C.; Lehner, J.R.; Maise, G.; Ng, K.K.; Mirsky, S.M.

    2002-01-01

    1 - Description of problem or function: CONTEMPT-4/MOD5 describes the response of multi-compartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user- supplied descriptions of compartments, inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. To accommodate degraded core type accidents, analytical models for hydrogen combustion within compartments and energy transfer due to gas radiation are also provided. CONTEMPT4/MOD6 is an update of previous CONTEMPT4 versions. Improvements in CONTEMPT4/MOD6 over CONTEMPT4/MOD3 include coding of a BWR pressure suppression system model, a hydrogen/carbon monoxide burn model, a gas radiation heat transfer model, a user specified variable junction (leakage) area as a function of pressure or time, additional heat transfer coefficient options for heat structures, generalized initial compartment conditions for inerted containment, an alternative containment spray model and spray carry-over capability. Also, the thermodynamic properties routines have been extended to accommodate the higher temperature and multicomponent gas mixtures associated with combustion. In addition, reduced running time is achieved by incorporation of an optional implicit numerical algorithm for junction flow. This makes economically feasible the analysis of very long

  5. Development and application of KEPRI realistic evaluation methodology (KREM) for LB-LOCA

    International Nuclear Information System (INIS)

    Ban, Chang-Hwan; Lee, Sang-Yong; Sung, Chang-Kyung

    2004-01-01

    A realistic evaluation method for LB-LOCA of a PWR, KREM, is developed and its applicability is confirmed to a 3-loop Westinghouse plant in Korea. The method uses a combined code of CONTEMPT4/MOD5 and a modified RELAP5/MOD3.1. RELAP5 code calculates system thermal hydraulics with the containment backpressure calculated by CONTEMPT4, exchanging the mass/energy release and backpressure in every time step of RELAP5. The method is developed strictly following the philosophy of CSAU with a few improvements and differences. Elements and steps of KREM are shown in Figure this paper. Three elements of CSAU are maintained and the first element has no differences. An additional step of 'Check of Experimental Data Covering (EDC)' is embedded in element 2 in order to confirm the validity of code uncertainty parameters before applying them to plant calculations. The main idea to develop the EDC is to extrapolate the code accuracy which is determined in step 8 to the uncertainties of plant calculations. EDC is described in detail elsewhere and the basic concepts are explained in the later section of this paper. KREM adopts nonparametric statistics to quantify the overall uncertainty of a LB-LOCA at 95% probability and 95% confidence level from 59 plant calculations according to Wilks formula. These 59 calculations are performed in step 12 using code parameters determined in steps 8 and 9 and operation parameters from step 11. Scale biases are also evaluated in this step using the information of step 10. Uncertainties of code models and operation conditions are reflected in 59 plant calculations as multipliers to relevant parameters in the code or as input values simply. This paper gives the explanation on the overall structures of KREM and emphasizes its unique features. In addition, its applicability is confirmed to a 3-loop plant in Korea. KREM is developed for the realistic evaluation of LB-LOCA and its applicability is successfully demonstrated for the 3-loop power plants in

  6. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    Science.gov (United States)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

  7. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    Energy Technology Data Exchange (ETDEWEB)

    Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

    2015-11-15

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  8. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    International Nuclear Information System (INIS)

    Urbonavičius, E.; Povilaitis, M.; Kontautas, A.

    2015-01-01

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  9. Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

    Energy Technology Data Exchange (ETDEWEB)

    De, T K; Collins, W M; Holmes, R W [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    CANDU nuclear reactors use D{sub 2}0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs.

  10. A simple analytical scaling method for a scaled-down test facility simulating SB-LOCAs in a passive PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il

    1992-02-01

    A Simple analytical scaling method is developed for a scaled-down test facility simulating SB-LOCAs in a passive PWR. The whole scenario of a SB-LOCA is divided into two phases on the basis of the pressure trend ; depressurization phase and pot-boiling phase. The pressure and the core mixture level are selected as the most critical parameters to be preserved between the prototype and the scaled-down model. In each phase the high important phenomena having the influence on the critical parameters are identified and the scaling parameters governing the high important phenomena are generated by the present method. To validate the model used, Marviken CFT and 336 rod bundle experiment are simulated. The models overpredict both the pressure and two phase mixture level, but it shows agreement at least qualitatively with experimental results. In order to validate whether the scaled-down model well represents the important phenomena, we simulate the nondimensional pressure response of a cold-leg 4-inch break transient for AP-600 and the scaled-down model. The results of the present method are in excellent agreement with those of AP-600. It can be concluded that the present method is suitable for scaling the test facility simulating SB-LOCAs in a passive PWR

  11. A through calculation of 1,100 MWe PWR large break LOCA by THYDE-P1 EM model

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

    THYDE-P1 is a code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs). Up to now, THYDE-P1 has been applied to various experiment analyses, which show its high capability to analyze LOCAs as a best estimate (BE) calculation code. In this report, evaluation model (EM) calculation method, especialy in the blowdown and refill phases, is established equivalently to WREM/J2 which is regarded as appropriate for an EM calculation code, and the results of them are compared and discussed. The present calculation was the first executed by THYDE-P1-EM, and was performed as Sample Calculation Run 80 which was a part of a series of THYDE-P sample calculations. The calculation was carried out from the LOCA initiation till 400 seconds for a guillotine break at the cold leg of a commercial 1,100 MWe PWR plant. The calculated results agreed well to that of the WREM/J2 code. (author)

  12. Prediction of CANDU-6 moderator system response following a large break LOCA using a 3D model

    International Nuclear Information System (INIS)

    De, T.K.; Collins, W.M.; Holmes, R.W.

    1995-01-01

    CANDU nuclear reactors use D 2 0 as a moderator inside the calandria vessel. Heat is generated in the calandria by neutron and gamma radiation from nuclear fission. During normal operating condition, hot moderator fluid is continuously pumped out from the bottom of the CANDU-6 calandria. After passing through a heat exchanger, the cooled moderator fluid is returned to the calandria through inlet nozzles. In the unlikely event of a loss of coolant accident (LOCA) the moderator acts as a heat sink. To predict moderator system response following a large break (reactor inlet header break) LOCA, a simulation was undertaken for Class IV power available (i.e. main moderator pump running) as well as for Class IV power unavailable during the LOCA. The analysis was performed to facilitate the assessment of fuel channel integrity following pressure tube (PT) and calandria tube (CT) contact by estimating the subcooling available during the inlet header break. The 3D code PHOENICS2 developed by CHAM U.K was used for the simulation. The results show an asymmetric flow pattern within the moderator both in the axial Z-direction of the calandria as well as in the X-Y plane. The temperature distribution within the moderator system shows, that hot spots are generated in areas, where the flow approaches stagnation. Hot spot temperatures are higher with Class IV power unavailable. (author). 1 ref., 12 figs

  13. Benchmark Tests to Develop Analytical Time-Temperature Limit for HANA-6 Cladding for Compliance with New LOCA Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Yong; Jang, Hun; Lim, Jea Young; Kim, Dae Il; Kim, Yoon Ho; Mok, Yong Kyoon [KEPCO Nuclear Fuel Co. Ltd., Daejeon (Korea, Republic of)

    2016-10-15

    According to 10CFR50.46c, two analytical time and temperature limits for breakaway oxidation and postquench ductility (PQD) should be determined by approved experimental procedure as described in NRC Regulatory Guide (RG) 1.222 and 1.223. According to RG 1.222 and 1.223, rigorous qualification requirements for test system are required, such as thermal and weight gain benchmarks. In order to meet these requirements, KEPCO NF has developed the new special facility to evaluate LOCA performance of zirconium alloy cladding. In this paper, qualification results for test facility and HT oxidation model for HANA-6 are summarized. The results of thermal benchmark tests of LOCA HT oxidation tester is summarized as follows. 1. The best estimate HT oxidation model of HANA- 6 was developed for the vender proprietary HT oxidation model. 2. In accordance with the RG 1.222 and 1.223, Benchmark tests were performed by using LOCA HT oxidation tester 3. The maximum axial and circumferential temperature difference are ± 9 .deg. C and ± 2 .deg. C at 1200 .deg. C, respectively. At the other temperature conditions, temperature difference is less than 1200 .deg. C result. Thermal benchmark test results meet the requirements of NRC RG 1.222 and 1.223.

  14. Notes on the Implementation of Non-Parametric Statistics within the Westinghouse Realistic Large Break LOCA Evaluation Model (ASTRUM)

    International Nuclear Information System (INIS)

    Frepoli, Cesare; Oriani, Luca

    2006-01-01

    In recent years, non-parametric or order statistics methods have been widely used to assess the impact of the uncertainties within Best-Estimate LOCA evaluation models. The bounding of the uncertainties is achieved with a direct Monte Carlo sampling of the uncertainty attributes, with the minimum trial number selected to 'stabilize' the estimation of the critical output values (peak cladding temperature (PCT), local maximum oxidation (LMO), and core-wide oxidation (CWO A non-parametric order statistics uncertainty analysis was recently implemented within the Westinghouse Realistic Large Break LOCA evaluation model, also referred to as 'Automated Statistical Treatment of Uncertainty Method' (ASTRUM). The implementation or interpretation of order statistics in safety analysis is not fully consistent within the industry. This has led to an extensive public debate among regulators and researchers which can be found in the open literature. The USNRC-approved Westinghouse method follows a rigorous implementation of the order statistics theory, which leads to the execution of 124 simulations within a Large Break LOCA analysis. This is a solid approach which guarantees that a bounding value (at 95% probability) of the 95 th percentile for each of the three 10 CFR 50.46 ECCS design acceptance criteria (PCT, LMO and CWO) is obtained. The objective of this paper is to provide additional insights on the ASTRUM statistical approach, with a more in-depth analysis of pros and cons of the order statistics and of the Westinghouse approach in the implementation of this statistical methodology. (authors)

  15. A methodology for the estimation of release of fission products during LOCA with loss of ECCS

    International Nuclear Information System (INIS)

    Lele, H.G.; Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K.; Venkat Raj, V.

    2002-01-01

    A Loss of Coolant Accident (LOCA) in a nuclear reactor along with the failure of the Emergency Core Cooling System can cause sustained voiding of the core. In such a situation the core experiences very low flow which leads to poor heat removal from the reactor core. The heat to be removed from the core includes stored heat, heat generated due to metal water reaction at high temperatures, decay heat etc. The poor heat removal leads to heating of the fuel pins to high temperatures. The heating of fuel pins is further enhanced due to metal-water reaction at high temperatures. These high temperatures of the fuel pins can lead to fission product release, which is transported into the Primary Heat Transport (PHT) system and can enter the containment through the break. Analysis is involved due to the complexity of the system and the phenomena to be simulated. In this paper a multistage analysis methodology is presented that involves the development and application of a number of computer programs to model the various phenomena involved. The computer code PHTACT computes the activity release from the fuel as a function of fuel temperatures and cladding oxidation, its distribution into the PHT system and release into the containment. Computation of thermal hydraulic parameters during LOCA is done using the thermal hydraulic analysis code RELAP5. The detailed simulation of fuel pin temperatures is done using computer code HT/MOD4. The convective boundary conditions required for the code are obtained from RELAP5. Creep deformation is considered in the computation of dimensional changes of the coolant channel and estimation of flow blockage due to clad ballooning. The progression of various reaction layers due to high temperature reaction between fuel and clad and clad and steam is also computed, which affects the structural strength of the clad. Different approaches are possible and analysis can be carried out in different phases depending upon the complexities to be

  16. Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA

    International Nuclear Information System (INIS)

    Cicero, G.M.; Briere, E.; Fornaciari, G.

    1994-06-01

    In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods

  17. RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

    2016-08-15

    Highlights: • RELAP5/SCDAPSIM model of AP1000 has been developed. • Analysis involves a LBLOCA (double ended guillotine break) study in cold leg. • Results are compared with those of WCOBRA–TRAC and TRACE. • Concluded that PCT does not violate the safety criteria of 1477 K. - Abstract: The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved

  18. Debris transport evaluation during the blow-down phase of a LOCA using computational fluid dynamics

    International Nuclear Information System (INIS)

    Park, Jong Pil; Jeong, Ji Hwan; Kim, Won Tae; Kim, Man Woong; Park, Ju Yeop

    2011-01-01

    Highlights: → We conducted CFD simulation on the spreading of the coolant in the containment after a break of the hot leg. It is used to estimate the dispersion of the debris within the containment. → It was assumed that the small and fine debris is transported by the discharge flow so that a fraction of the small and fine debris transport can be estimated based on the amount of water. → The break flow was assumed to be a homogeneous two-phase mixture without phase separation. Isenthalpic expansion of the break flow was used to specify the inlet boundary condition of the break flow. → The fraction of the small and fine debris transported to the upper part is 73%; this value is close to the value calculated using 1D lumped-parameter codes by the USNRC and the KINS, respectively, while 48% more than the value shown in the NEI 04-07. - Abstract: The performance of the emergency recirculation water sump under the influence of debris accumulation following a loss-of-coolant accident (LOCA) has long been of safety concern. Debris generation and transport during a LOCA are significantly influenced by the characteristics of the ejected coolant flow. One-dimensional analyses previously have been attempted to evaluate the debris transport during the blow-down phase but the transport evaluation still has large uncertainties. In this work, a computational fluid dynamics (CFD) analysis was utilized to evaluate small and fine debris transport during the blow-down phase of a pressurized water reactor, OPR1000. The coolant ejected from the ruptured hot-leg was assumed to expand in an isenthalpic process. The transport of small and fine debris was assumed to be dominated by water-borne transport, and the transport fractions for the upper and lower parts of the containment were quantified based on the CFD analysis. It was estimated that 73% of small and fine debris is transported to the upper part of the containment. This value is close to the values estimated by nuclear

  19. Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

    International Nuclear Information System (INIS)

    Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

    2004-01-01

    The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

  20. Evaluation of the pressure difference across the core during PWR-LOCA reflood phase

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Murao, Yoshio

    1979-03-01

    The flooding rate of the core influences largely cooling of the core during the reflood phase of a PWR-LOCA. Since the void fraction of two-phase flow in the core is important determining the flooding rate, it is essential to examine this void fraction. The void fraction in the core during the reflood phase obtained by experiment was compared with those predicted by the correlations respectively of Akagawa, Nicklin, Zuber, Yeh, Griffice, Behringer and Jhonson. Only Yeh's correlation was found to be usable for the purpose. The pressure difference of the core during the reflood phase was calculated by reflood analyzing code REFLA-1D using Yeh's correlation. Following are the results: (1) During the steady-state period after quenching of the heaters, the prediction agrees within +-15% with the experiment. (2) During the transient period when the quench front is advancing, the prediction is not in agreement with the experiment, the difference being about +-40%. Influence of the advancing quench front upon the void fraction in the core must further be studied. (author)

  1. Applicability of TASS/SMR using drift flux model for SMART LOCA analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young-Jong, E-mail: chung@kaeri.re.kr; Kim, Soo Hyung; Lee, Gyu Hyeung; Lee, Won Jae

    2013-09-15

    Highlights: • SMART plant and TASS/SMR code have been developed by KAERI. • TASS/SMR code adopts a drift flux model to consider relative velocity under two-phase condition. • Drift flux model in TASS/SMR is validated using separate effect test results. • Applicability of TASS/SMR using drift flux model for SMART LOCA analysis is confirmed. -- Abstract: Small reactors can apply to local power demands or remote areas. SMART, which can produce 90 MWe of electricity and 40,000 tons/day of sea-water desalination for a 100,000 population city, is a promising advanced integral type small reactor. The thermal hydraulic analysis code with a drift flux model, TASS/SMR, was developed for a conservative simulation of a small break loss of coolant accident in SMART. Taking into account SMART-specific inherent characteristics, the code adopts the Chexal–Lellouche correlation for a drift flux model. The capability of TASS/SMR code is validated using the results of the experimental data. The code predicts conservatively the void distribution compared with the experimental data. TASS/SMR calculation predicts reasonably major phenomena for the SBLOCA and is more conservative than or nearly the same as the results of the best estimated realistic system code, MARS. TASS/SMR code can be used for a SBLOCA analysis of SMART.

  2. RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

    International Nuclear Information System (INIS)

    Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

    1984-01-01

    The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

  3. Comparative analysis of a LOCA for a German PWR with ASTEC and ATHLET-CD

    International Nuclear Information System (INIS)

    Reinke, N.; Chan, H.W.; Sonnenkalb, M.

    2013-01-01

    This paper presents the results of a comparative analysis performed with ASTEC V2.02 and a coupled ATHLET-CD V2.2c /COCOSYS V2.4 calculation for a German 1300 MWe KONVOI type PWR. The purpose of this analysis is mainly to assess the ASTEC code behaviour in modelling of both the thermal-hydraulic phenomena in the coolant circuit arising during a hypothetical severe accident and the early phase of the core degradation versus the more mechanistic code system ATHLET-CD/COCOSYS. The performed analyses cover a loss of coolant accident sequence (LOCA). Such comparison has been done for the first time. The integral code ASTEC (Accident Source Term Evaluation Code) commonly developed since 1996 by IRSN and GRS is a fast running programme, which allows the calculation of entire sequences of severe accidents (SA) in light water reactors from the initiating event up to the release of fission products into the environment, thereby covering all important in-vessel and containment phenomena. The thermal-hydraulic mechanistic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by GRS for the analysis of the whole spectrum of leaks and transients in PWRs and BWRs. For modeling of core degradation processes the CD part (Core Degradation) of ATHLET can be activated. For analyses of the containment behavior, ATHLET-CD has been coupled to the GRS code COCOSYS (COntainment COde SYStem). (orig.)

  4. NEPTUN/5052, PWR LOCA Cooling Heat Transfer Tests for Loft, Reflood Test

    International Nuclear Information System (INIS)

    Richner, M.; Analytis, G.Th.; Aksan, S.N.

    1993-01-01

    1 - Description of test facility: NEPTUN is designed to perform PWR LOCA simulation experiments, which provide the full length emergency cooling heat transfer tests for LOFT. Therefore the NEPTUN heater bundle with 33 electrical heater elements and 4 guide tubes simulates a section of the LOFT nuclear core. The main test loop also contains measuring systems for the carry-over rate and for the steam expelled, and a back-pressure control system. A water loop brings the water to the initial reflooding conditions. In addition, auxiliary systems maintain normal operating conditions. 2 - Description of test: Test 5052 is one of a series of 40 reflood tests performed in NEPTUN. Before the start of the test, the flooding water in its circuit is brought to the following conditions: pressure = 4.1 bar; velocity = 2.5 cm/sec; subcooling temperature = 78 C; single rod power = 2.45 kW; maximal initial cladding temperature = 867 C. 3 - Status: CSNI1013/01, 21-Jul-1993 Arrived at NEADB

  5. TWODEE-2/MOD3, 2-D Time-Dependent Fuel Elements Thermal Analysis after PWR LOCA

    International Nuclear Information System (INIS)

    Lauben, G. N.

    2001-01-01

    1 - Description of problem or function: WREM-TOODEE2 is a two- dimensional, time-dependent, fuel-element thermal analysis program. Its primary purpose is to evaluate fuel-element thermal response during post-LOCA refill and reflood in a pressurized water reactor (PWR). 2 - Method of solution: TOODEE2 calculations are carried out in a two-dimensional mesh region defined in slab or cylindrical geometry by orthogonal grid lines. Coordinates which form order pairs are labeled x-y in slab geometry, and those in cylindrical geometry are labeled r-z for the axisymmetric case and r-theta for the polar case. Conduction and radiation are the only heat transfer mechanisms assumed within the boundaries of the mesh region. Convective and boiling heat transfer mechanisms are assumed at the boundaries. The program numerically solves the two-dimensional, time-dependent, heat conduction equation within the mesh region. 3 - Restrictions on the complexity of the problem: WREM-TOODEE2 considers only axisymmetric geometry although the equations for slab and polar geometry are included in the program

  6. Applicability of TASS/SMR using drift flux model for SMART LOCA analysis

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Kim, Soo Hyung; Lee, Gyu Hyeung; Lee, Won Jae

    2013-01-01

    Highlights: • SMART plant and TASS/SMR code have been developed by KAERI. • TASS/SMR code adopts a drift flux model to consider relative velocity under two-phase condition. • Drift flux model in TASS/SMR is validated using separate effect test results. • Applicability of TASS/SMR using drift flux model for SMART LOCA analysis is confirmed. -- Abstract: Small reactors can apply to local power demands or remote areas. SMART, which can produce 90 MWe of electricity and 40,000 tons/day of sea-water desalination for a 100,000 population city, is a promising advanced integral type small reactor. The thermal hydraulic analysis code with a drift flux model, TASS/SMR, was developed for a conservative simulation of a small break loss of coolant accident in SMART. Taking into account SMART-specific inherent characteristics, the code adopts the Chexal–Lellouche correlation for a drift flux model. The capability of TASS/SMR code is validated using the results of the experimental data. The code predicts conservatively the void distribution compared with the experimental data. TASS/SMR calculation predicts reasonably major phenomena for the SBLOCA and is more conservative than or nearly the same as the results of the best estimated realistic system code, MARS. TASS/SMR code can be used for a SBLOCA analysis of SMART

  7. Experiment of the downcomer effective water head during a reflood phase of PWR LOCA

    International Nuclear Information System (INIS)

    Sudo, Yukio; Murao, Yoshio

    1978-12-01

    The results and analysis are described of a downcomer effective water head experiment. Downcomer effective water head is the driving force to feed an emergency coolant to the core during a reflood phase of PWR LOCA. The test rig has dimensions of the full-scale height and gap. Experimental conditions are: downcomer wall temperature = 250 0 -- 300 0 C, back pressure = 1 atm, coolant temperature = 98 0 -- 100 0 C, extraction water velocity = 0 -- 2 cm/s, and gap size = 200 mm. The effective water head histories obtained by experiment were compared with those predicted from the heat release from the downcomer walls. The heat release was calculated from the temperature histories indicated by thermocouples instrumented in and on the walls during experiment. The following were revealed: (1) The relation of heat flux and superheat (q vs ΔT sub(s)) obtained in the experiment is much different from that in pool boiling. (2) The predicted effective water head is in good agreement with the experimental one after 120 sec from the initiation of coolant injection. (3) The effect of extraction water velocity is negligible. (4) The effect of initial wall temperatures is evident. (author)

  8. Probabilistic Dose Assessment from SB-LOCA Accident in Ujung Lemahabang Using TMI-2 Source Term

    Directory of Open Access Journals (Sweden)

    Sunarko

    2017-01-01

    Full Text Available Probabilistic dose assessment and mapping for nuclear accident condition are performed for Ujung Lemahabang site in Muria Peninsula region in Indonesia. Source term is obtained from Three-Mile Island unit 2 (TMI-2 PWR-type SB-LOCA reactor accident inverse modeling. Effluent consisted of Xe-133, Kr-88, I-131, and Cs-137 released from a 50 m stack. Lagrangian Particle Dispersion Method (LPDM and 3-dimensional mass-consistent wind field are employed to obtain surface-level time-integrated air concentration and spatial distribution of ground-level total dose in dry condition. Site-specific meteorological data is obtained from hourly records obtained during the Site Feasibility Study period in Ujung Lemahabang. Effluent is released from a height of 50 meters in uniform rate during a 6-hour period and the dose is integrated during this period in a neutrally stable atmospheric condition. Maximum dose noted is below regulatory limit of 1 mSv and radioactive plume is spread mostly to the W-SW inland and to N-NE from the proposed plant to Java Sea. This paper has demonstrated for the first time a probabilistic analysis method for assessing possible spatial dose distribution, a hypothetical release, and a set of meteorological data for Ujung Lemahabang region.

  9. Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

    International Nuclear Information System (INIS)

    Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

    2010-01-01

    SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

  10. Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Isozaki, Toshikuni; Yano, Toshikazu; Miyazaki, Noriyuki; Kato, Rokuro; Kurihara, Ryoichi; Ueda, Shuzo; Miyazono, Shohachiro

    1982-09-01

    Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

  11. Small break LOCA analysis for YGN 5 and 6 RCP trip strategy in power mode operation

    International Nuclear Information System (INIS)

    Kim, Tech Mo; Choi, Han Rim

    2001-01-01

    A continued operation of Reactor Coolant Pumps(RCPs) during a Small Break Loss of Coolant Accident(SBLOCA) in all operation mode may increase unnecessary inventory loss from the Reactor Coolant System(RCS) causing a severe core uncovery which might lead to fuel failure. After Three Mile Island Unit 2(TMI-2) accident, the Combustion Engineering Owner Group(CEOG) developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2). The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis demonstrates the inherent safety of RCP trip strategy during an SBLOCA for Youggwang Nuclear Power Plant Unit 5 and 6(YGN 5 and 6). The trip setpoint of the first two RCPs for YGN 5 and 6 is calculated to be 1721 psia in pressurizer pressure based on the limiting SBLOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 5 and 6 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at the worst time of minimum liquid inventory

  12. Analytical methods of leakage rate estimation from a containment under a LOCA

    International Nuclear Information System (INIS)

    Chun, M.H.

    1981-01-01

    Three most outstanding maximum flow rate formulas are identified from many existing models. Outlines of the three limiting mass flow rate models are given along with computational procedures to estimate approximate amount of fission products released from a containment to environment for a given characteristic hole size for containment-isolation failure and containment pressure and temperature under a loss of coolant accident. Sample calculations are performed using the critical ideal gas flow rate model and the Moody's graphs for the maximum two-phase flow rates, and the results are compared with the values obtained from then mass leakage rate formula of CONTEMPT-LT code for converging nozzle and sonic flow. It is shown that the critical ideal gas flow rate formula gives almost comparable results as one can obtain from the Moody's model. It is also found that a more conservative approach to estimate leakage rate from a containment under a LOCA is to use the maximum ideal gas flow rate equation rather than the mass leakage rate formula of CONTEMPT-LT. (author)

  13. Design of Test Facility to Evaluate Boric Acid Precipitation Following a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong-Kwan; Song, Yong-Jae [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The U.S.NRC has identified a concern that debris associated with generic safety issue (GSI) - 191 may affect the potential precipitation of boric acid due to one or more of the following phenomena: - Reducing mass transport (i.e. mixing) between the core and the lower plenum (should debris accumulate at the core inlet) - Reduced lower plenum volume (should debris settle in the lower plenum), and, - Increased potential for boric acid precipitation (BAP) in the core (should debris accumulate in suspension in the core) To address these BAP issues, KHNP is planning to conduct validation tests by constructing a BAP test facility. This paper describes the design of test facility to evaluate BAP following a LOCA. The design of BAP test facility has been developed by KHNP. To design the test facility, test requirements and success criteria were established, and scaling analysis of power-to-volume method, Ishii-Kataoka method, and hierarchical two-tiered method were investigated. The test section is composed of two fuel assemblies with half of full of prototypic FA height. All the fuel rods are heated by the electric power supplier. The BAP tests in the presence of debris, buffering agents, and boron will be performed following the test matrix.

  14. Experimental study of effect of initial clad temperature on reflood phenomena during PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun; Murao, Yoshio

    1983-01-01

    Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWR-LOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo-hydraulics under the simulated core inlet flow conditions. However, the calculated temperature rise of the maximum powered rod based on the one-dimensional core analysis was higher than that of the average powered rod, which contradicts the tendency observed in CCTF tests. (author)

  15. Pipe rupture test results; 6 in. pipe whip test under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi; Yano, Toshikazu; Ueda, Shuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kato, Rokuro; Miyazono, Shohachiro

    1983-02-01

    A series of pipe rupture tests has been performed in JAERI to demonstrate the safety of the primary coolant circuits in the event of pipe rupture, in nuclear power plants. The present report summarizes the results of 6 in. pipe whip tests (RUN 5605, 5606), under BWR LOCA conditions (285 0 C, 6.8 MPa), which were performed in August, 1981. The test pipe is made of Type 304 stainless steel and its outer diameter is 6 in. and its thickness is 11.1 mm. The restraints are made of Type 304 stainless steel and its diameter is 16.0 mm. Two restraints were set on the restraint support with clearance of 100 mm. Overhang length was varied as the parameter in these tests and was 300 mm or 700 mm. The following results are obtained. (1) The deformations of a pipe and restraints are limited effectively by shorter overhang length of 300. However, they become larger when the overhang length is 700 mm, and the pipe deforms especially at the setting point of restraints. (2) Velocity at the free end of pipe becomes about 30 m/sec just after the break. However, velocity at the setting point of restraint becomes about only 4 m/sec just after the break. (3) It seems from the comparison between the 4 in. tests and 6 in. tests that the maximum restraint force of 6 in. tests is about two times as large as that of 4 in. tests. (author)

  16. Analysis of cold leg LOCA with failed HPSI by means of integrated safety assessment methodology

    International Nuclear Information System (INIS)

    Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayorga, J.

    2014-01-01

    Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties

  17. Influence of pre-hydriding on embrittlement of E110 alloy under LOCA conditions

    International Nuclear Information System (INIS)

    VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Fedotov, P.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Kuznetsov, V.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Nechaeva, O.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Novikov, V.; VNIINM, Moscow (Russian Federation))" data-affiliation=" (SC VNIINM, Moscow (Russian Federation))" >Salatov, A.; Ignatiev, D.; Mokrushin, A.; Soldatkin, D.; Urusov, A.

    2015-01-01

    The researches presented in this paper were carried out in the framework of TVS-K project developed by JSC “TVEL”. The data on the corrosion and residual ductility of unirradiated and pre-hydrided E110 alloy under LACA conditions at temperature range from 1100 to 1200°C are presented. The hydrogen concentration was varied from 30 (as-received) to 600 wppm. The initial concentration of hydrogen has no effect on the oxidation kinetics, while the oxidation kinetics are parabolic and the breakaway oxidation is not observed. Oxide films on surfaces of claddings are black and shining. There are no cracks, visual spots and peelings. The residual ductility of oxidised samples decrease with hydrogen concentration rise. The residual ductility of claddings oxidized at 1100 °C, generally higher than the same of the claddings oxidized at 1200 °C. E110 alloy has a good residual ductility in comparison to Zry-4, ZIRLO, M5. Joint analysis of the test results allowed us to formulate embrittlement criteria of the E110 alloy under LOCA conditions. This embrittlement criterion is preliminary, because the experimental data base must to be enlarged by results of tests with claddings of another geometry and quench experiments. (author)

  18. A 25% double-ended LOCA in the PSB-VVER facility

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Lipatov, I.A.; Dremin, G.I.

    2003-01-01

    The paper presents the results of the experimental investigation in the PSB-VVER facility and post-test analysis with RELAP5/MOD3.3 of the thermal-hydraulic response of the PSB-VVER to a 25% double-ended Hot Leg Break (HLB). The test scenario included loss of off-site power concurrently with the scram-signal and the safety system operation as described in the reference VVER-1000 operational manual in the case of this type of accident assuming one diesel-generator failure. The key transient parameter trends as well as sequence of events and phenomena are given in the paper. RELAP5/MOD3.3 a post-test analysis has been performed using the experimental data gained as a base. The reasonable qualitative agreement between the key calculated and measured variables has been shown. The quantitative code accuracy evaluation has shown that the total average amplitude of the main parameters' deviations AA tot tot < 0.28 that corresponds to satisfactory quality of the VVER-1000 hot leg guillotine break LOCA modeling in the PSB-VVER. (author)

  19. Upper plenum break LOCA investigation in the ISB-VVER and PSB-VVER facilities

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V. N.; Melikhov, O. I.; Lipatov, I. A.; Nikonov, S. M.; Dremin, G. I.; Galchanskaya, S. A.; Gashenko, M. P.; Rovnov, A. A.; Kapustin, A. V.; Elkin, I. V. [EREC, Moscow (Russian Federation)

    2003-07-01

    The capability to define the actual NPP transient/accident scenario depends to a great extent on facilities' scaling and reliability of the system thermalhydraulic codes which, in turn, are assessed against the experimental data taken in the same facilities. At the present time, it is received fact that the rigorous modeling of the cumulative set of all thermalhydraulic processes in the plant primary and secondary sides during accident is unfeasible. Therefore, the extrapolation of the facilities loops behavior to the actual systems constitutes a fundamental problem in this area. In the paper, some aspects for the problem have been discussed in the course of comparative analysis of the data derived from the 11 % upper plenum break LOCA tests performed in the PSB-VVER and ISB-VVER integral test facilities under the close scenarios. Both facilities, PSB-VVER and ISB-VVER, are modeled the same VVER-1000 reactor in different scales. The thermalhydraulic behavior of the primary systems in both facilities has been discussed in the paper, and shown to be similar. Also, the attention has been focused upon the discrepancies in the significant variables trends. The discrepancies are shown to be caused by influence of peculiarities of the facilities hardware and due to the scale factor. The scaling study is an important aspect of the thermalhydraulic codes verification procedure. Being qualified against the experimentally simulated accident sequence in two test facilities of different scales, the thermalhydraulic codes will be capable of evaluation of the prototype behavior to greater accuracy.

  20. A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

    Directory of Open Access Journals (Sweden)

    Jae-Yong Kim

    2014-01-01

    Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

  1. Locas al Rescate: The Transnational Hauntings of Queer Cubanidad

    Directory of Open Access Journals (Sweden)

    Lázaro Lima

    2011-12-01

    Full Text Available

    Locas al Rescate: The Transnational Hauntings of Queer Cubanidad” (originally published in Cuba Transnational offers a significant contribution both to transnational American Studies and to gender studies. In telling the insider story of the alternative identity formation, practices, and forms of “rescue” initiated by the affective activism of the Cuban American society in drag in 1990s Miami/South Beach, Lima resuscitates the liberatory gestures of a subculture defined by its pursuit of its own acceptance, value, and freedom. With their aesthetic and political life on a raft, the gay micro-communities inside Cuban America asserted their own islandic space, Lima observes, performing “takeovers” in and of parks and bars and beaches—creating a post-Habermasian sphere of public activism focused on private parts, saving themselves from AIDS, from the disaffection and disaffiliation of the right-wing Cuban immigrant community, and from the failure of their own yearning to belong, to be wanted, to be embodied as the figure of their compelling Cubanidad. Against the hegemony of the invented collective politics of the sacrificing immigrants whose recognition of the queer side of being (of a being constituted by identity loss is yet to come, Lima suggests a spectral return—a personal and transnational reckoning of those whose lives the dream of freedom drowned.

  2. Pipe rupture test results: 4-inch pipe whip tests under PWR LOCA conditions

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki; Ueda, Shuzo; Isozaki, Toshikuni; Kato, Rokuro; Kurihara, Ryoichi; Yano, Toshikazu; Miyazono, Shohachiro

    1982-09-01

    This report summarizes the results of 4-inch pipe whip tests (RUN No. 5506, 5507, 5508 and 5604) under the PWR LOCA conditions. The dynamic behaviors of the test pipe and restraints were studied in the tests. In the tests, the gap between the test pipe and the restraints was kept at the constant value of 8.85 mm and the overhang length was varied from 250 mm to 650 mm. The dynamic behaviors of the test pipe and the restraint were made clear by the outputs of strain gages and the measurements of residual deformations. The data of water hammer in subcooled water were also obtained by the pressure transducers mounted on the test pipe. The main conclusions obtained from the tests are as follows. (1) The whipping of pipe can be prevented more effectively as the overhang length becomes shorter. (2) The load acting on the restraint-support structure becomes larger as the overhang length becomes shorter. (3) The restraint farther from the break location does not limit the pipe movement except for the first impact when the overhang length is long. (4) The ultimate moment M sub(u) of the pipe at the restraint location can be used to predict the plastic collapse of the whipping pipe. (5) The restraints slide along the pipe axis and are subjected to bending moment, when the overhang length is long. (author)

  3. Effects of post-LOCA conditions on a protective coating (paint) for the Nuclear Power Industry

    International Nuclear Information System (INIS)

    Loyola, V.M.; Womelsduff, J.E.

    1985-03-01

    When corrosion protection of steel cannot be achieved by galvanizing due to size, use, or other restrictions, the steel is frequently protected by the application of a suitable corrosion-inhibiting paint. A widely accepted corrosion inhibiting coating is one in which finely powdered zinc metal is dispersed in an organic polymer matrix and applied to steel as a paint. This system is often used with a non-zinc bearing topcoat for enhanced protection. We have studied the oxidation of zinc in a zinc-rich coating used in the nuclear power industry and have measured the rates of hydrogen generation from these coatings due to zinc oxidation at temperatures of up to 175 0 C. The results suggest that the real-time rates of hydrogen generation are considerably higher than previously believed. A second concern involves the generation of debris or solid reaction products which could cause plugging or fouling of the recirculation pumps, spray nozzles, and/or heat exchangers. Coatings are observed to fail at post-LOCA conditions which are well within the limits predicted by Design Basis Accident analysis. The failures involve cracking and/or delamination of the topcoat and production of solid corrosion products involving the zinc-rich primer. 22 refs., 10 figs., 6 tabs

  4. Improvement on models associated with LOCA and loss of RHR accidents during shutdown

    International Nuclear Information System (INIS)

    Chang, W. P.; Chung, Y. J.; Kim, W. S.; Kim, K. D.; Lee, S. J.; Jung, J. J.; Ha, G. S.; Son, Y. S.; Chung, B. D.; Han, D. H.; Lee, Y. J.; Hwang, T. S.; Lee, S. Y.; Park, C. Y.; Choi, H. R.; Lee, S. Y.; Choi, J. H.; Ban, C. H.; Bae, G. H.

    1997-07-01

    The characteristics of the best estimate codes available in Korea have been studied through literature surveys for the reliability on LOCA analyses and then, a feasibility study on reduction of capacities of existing safety systems in YGN 3/4 have been carried out using the codes. Since it has been expected to adopt DVI + 4 -Train HPSI in the next generation reactor, the core uncoveries under one DVI line break and 6 cold leg break, which is a requirement for advance d reactor by EPRI, in addition to LBLOCA for reduction effect of SIT capacity, have been analyzed. Finally, an effort on finding the way how the system could be simplified, has been made through the analysis of SIT injection characteristics. On the other hand, the best estimate methodology consisting of uncertainties of the code itself, bias, and application have been developed first and quantification of the uncertainty has been made the case of KORI unit 3 afterward. The prediction capabilities of the best estimate codes and major physical models on the accident under loss of RHR during shutdown have been assessed suing the large scale experimental data delivered from France and then, the assessed codes have been used to provide essential data required for description of operation procedures in YGN 3/4. (author). 64 refs., 45 figs

  5. MAAP-CANDU simulations of LOCA/LOECI accidents at Darlington NGS

    International Nuclear Information System (INIS)

    Kwee, M.T.; Choi, M.H.; Leung, R.K.

    1996-01-01

    Severe accidents have been the subject of a great deal of analysis and research, particularly in the light water reactor community. Although severe accident analysis in Canada deuterium-uranium (CANDU) reactors has not been published abundantly, a significant body of research and analysis has been accumulated. This has occurred because CANDU has directly taken into consideration a set of severe accidents [e.g loss-of-coolant accidents (LOCAs) coincident with a loss-of-emergency-coolant injection (LOECI)] in the design basis. These accidents have served to define the design requirements that ensure the integrity of the heat transport system. The CANDU reactor design has inherent heat sinks such as the primary heat transport system, the secondary side, moderator system, and shielding system (shield tank and end shields). These heat sinks are significant and are able to moderate or terminate the progression of severe accidents that go beyond the design base cases. These types of accidents are typically analyzed at Ontario Hydro in conjunction with probabilistic safety analysis (PSA), where the severe accident consequences are analyzed by a series of conservative hand-calculation methods

  6. Application of thermal hydraulic and severe accident code SOCRAT/V3 to bottom water reflood experiment QUENCH-LOCA-0

    International Nuclear Information System (INIS)

    Vasiliev, A.D.; Stuckert, J.

    2013-01-01

    Highlights: ► QLOCA-0 test simulates a design basis LOCA NPP accident with maximum temperature 1300 K. ► Deep understanding of hydraulics and thermal mechanics under accident conditions is necessary. ► We model the test QLOCA-0 with bottom flooding using the Russian code SOCRAT/V3. ► Calculated and experimental data are in a good agreement. ► Experimental procedure is determined to reach a representative LOCA scenario in future tests. -- Abstract: The thermal hydraulic and SFD (severe fuel damage) best estimate computer modeling code SOCRAT/V3 has been used for the calculation of QUENCH-LOCA-0 experiment. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario of the LOCA (loss of coolant accident) nuclear power plant accident sequence in which the overheated up to 1300 K reactor core would be reflooded from the bottom by ECCS (emergency core cooling system). The first test QUENCH-LOCA-0 was successfully conducted at the KIT, Karlsruhe, Germany, in July 22, 2010, and was performed as the commissioning test for this series. The rod claddings are identical to that used in PWRs. The bundle was electrically heated in steam from 800 K to 1340 K with the heat-up rate of approximately 2.7 K/s. After cooling in the saturated steam the bottom flooding with water flow rate of about 100 g/s was initiated. The SOCRAT calculated results are in a good agreement with experimental data taking into account additional quenching due to water condensate entrainment at the steam cooling stage. SOCRAT/V3 has been used for estimation of further steps in experimental procedure to reach a representative LOCA scenario in future tests

  7. Input data preparation and simulation of the second standard problem of IAEA using the Trac/PF1 code

    International Nuclear Information System (INIS)

    Madeira, A.A.; Pontedeiro, A.C.; Silva Galetti, M.R. da; Borges, R.C.

    1989-10-01

    The second Standard Problem sponsored by IAEA consists in the simulation of a small LOCA located in the downcomer of a PMK-NVH integral test facility, which models WWER/440 type reactor. This report presents input data preparation and comparison between TRAC-PF1 results and experimental measurements. (author) [pt

  8. Establishment of the operating procedure to prevent boron precipitation during Post-LOCA long term cooling for Korean Westinghouse 3-loop NPPs

    International Nuclear Information System (INIS)

    Choi, Han Rim; Kwon, Tae Soon; Ban, Chang Hwan; Jeong, Jae Hoon; Lee, Young Jin.

    1996-11-01

    During post-LOCA LTC the increase of the excess reactivity for the extended fuel cycle should require increasing the RWST boron concentration in order to ensure core subcritical state. To quantify the concentration increment, the calculation methods was developed for the post-LOCA RCS/Sump mixed mean boron concentration, which applied for Kori 3 and 4 and Ulchin 1 and 2 of the Westinghouse 3-loop nuclear power plants in Korean. From the calculation results, the minimum boric acid concentrations increased of the RWST and accumulator were determined consideration of the convenient operation for operator on reloading. Boric acid concentrations of the RWST and the accumulators for Westinghouse 3-loop type plants were increased to meet the post-LOCA shutdown requirement for the long life cycles from 12 months to 18 months. To maintain LTC capability following a LOCA, the switchover time is examined using boron code of prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results showed that hot leg recirculation switchover times were shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3 and 4 and 8 hours from 18 hours for Ulchin 1 and 2, respectively. The flow path in the mode J for Kori 3 and 4 was recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1 and 2. (author). 2 tabs., 12 figs., 13 refs

  9. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Directory of Open Access Journals (Sweden)

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  10. Scientific design of the test facility for the KNGR DVI line small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Park, Choon Kyung; Jun, Hyung Gil; Cho, Seok; Kwon, Tae Soon; Song, Chul Hwa; Kim, Jung Taek

    1999-03-01

    Scientific design of the experimental facility (OASIS) for the KNGR (Korea Next Generation Reactor) DVI line SB-LOCA simulation is carried out. Main purpose of the OASIS is to produce thermal-hydraulic data base for determining the best location of the DVI (Direct Vessel Injection) injection nozzle of the KNGR as well as verifying its design performance in view of the ECCS (Emergency Core Cooling System) effectiveness. The experimental facility is designed based on the Ishii's three-level scaling law. The facility has 1/4 height and 1/341 area scaling ratio. It corresponds to the volume scale of 1/1364. The power scaling is 1/682 and the system pressure is prototypic. The OASIS consists of a core, a downcomer, two steam generators, two pump simulators, a break simulator, a collection tank, primary piping as well as a circulation pump for initial test condition. Each component is designed based on the Ishill's global scaling and boundary flow scaling of mass, energy and momentum. In addition, local phenomena scaling is carried out for the design of major components to preserve key local phenomena in each component. Most of the key phenomena are well preserved in the OASIS. However, the local scaling analysis shows that distortions of the void fraction and mixture level can not be avoided in the core. It comes from the basic features of the Ishill's scaling law in case of the reduced-height simulation. However, it is expected that these distortions will be analyzed properly by a best estimate system analysis code. (Author). 22 refs., 20 tabs., 25 figs.

  11. Analysis of insulation material deterioration under the LOCA simulated environment on the basis of reaction kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Okada, Sohei; Kusama, Yasuo; Ito, Masayuki; Yagi, Toshiaki; Yoshikawa, Masato (Japan Atomic Energy Research Inst., Takasaki, Gunma. Takasaki Radiation Chemistry Research Establishment)

    1982-12-01

    In the type test of the electric cables installed in reactor containment vessels, it is considerably difficult to perform the testing over a year once in a while to simulate the accidental environment containing radiation and high temperature steam. Two requirements which seem to be more realistic as compared with the above mentioned testing method are inconsistent with each other. To solve this problem, a general rule of deterioration or the expression by an equation is necessary, which enables the extrapolation to show that a short term testing stands on the safety side. The authors have tried to numerically analyze the change of mechanical characteristics of ethylene-propylene rubber (EPR) and Hypalon which are, important as the materials for PH cables (fire-retardant, EP rubber-insulated, chlorosulfonated polyethylene-sheathed cable), in a complex environment of radiation, steam and chemical spray simulating PWR LOCA conditions. In this report, a method is proposed to analyze and estimate the properties by the regression analysis technique on the basis of reaction kinetics, and the analyzed results are described in the order of experiment, analysis method and the results and consideration. The deterioration of the elongation P = e/esub(o) of EPR and Hypalon in the above described complex environment can be represented by the equation - dP/dt = KPsup(n). The exponent n varied in the cases when air is contained or not in that environment, suggesting that the different reactions are dominant in both conditions, respectively. For EPR, n was close to 2 if air was not contained and close to 1 if air was contained in the system.

  12. Mixing of radiolytic hydrogen generated within a containment compartment following a LOCA

    International Nuclear Information System (INIS)

    Willcutt, G.J.E. Jr.; Gido, R.G.

    1978-07-01

    The objective of this work was to determine hydrogen concentration variations with position and time in a closed containment compartment with radiolytic hydrogen generation in the water on the compartment floor following a Loss-of-Coolant-Accident (LOCA). One application is to determine the potential difference between the compartment maximum hydrogen concentration and a hydrogen detector reading, due to the detector location. Three possible mechanisms for hydrogen transport in the compartment were investigated: (1) molecular diffusion, (2) possible bubble formation and motion, and (3) natural convection flows. A base case cubic compartment with 6.55-m (21.5-ft) height was analyzed. Parameter studies were used to determine the sensitivity of results to compartment size, hydrogen generation rates, diffusion coefficients, and the temperature difference between the floor and the ceiling and walls of the compartment. Diffusion modeling indicates that if no other mixing mechanism is present for the base case, the maximum hydrogen volume percent (vol percent) concentration difference between the compartment floor and ceiling will be 4.8 percent. It will be 24.5 days before the maximum concentration difference is less than 0.5 percent. Bubbles do not appear to be a potential source of hydrogen pocketing in a containment compartment. Compartment natural convection circulation rates for a 2.8 K (5 0 F) temperature difference between the floor and the ceiling and walls are estimated to be at least the equivalent of 1 compartment volume per hour and probably in the range of 4 to 9 compartment volumes per hour. Related natural convection studies indicate there will be turbulent mixing in the compartment for a 2.8 K (5 0 F) temperature difference between the floor and the ceiling and walls

  13. Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

    International Nuclear Information System (INIS)

    Papini, Davide; Grgic, Davor; Cammi, Antonio; Ricotti, Marco E.

    2011-01-01

    Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes). The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.

  14. Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1989-01-01

    This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

  15. Realistic methods for calculating the releases and consequences of a large LOCA

    International Nuclear Information System (INIS)

    Stephenson, W.; Dutton, L.M.C.; Handy, B.J.; Smedley, C.

    1992-01-01

    This report describes a calculational route to predict realistic radiological consequences for a successfully terminated large-loss-of-coolant accident (LOCA) at a pressurized-water reactor (PWR). All steps in the calculational route are considered. For each one, a brief comment is made on the significant differences between the methods of calculation that were identified in the benchmark studies and recommendations are made for the methods and data for carrying out realistic calculations. These are based on the best supportable methods and data and the technical basis for each recommendation is given. Where the lack of well-validated methods or data means that the most realistic method that can be justified is considered to be very conservative, the need for further research is identified. The behaviour of inorganic iodine and the removal of aerosols from the atmosphere of the reactor building are identified as areas of particular importance. Where the retention of radioactivity is sensitive to design features, these are identified and, for the most importance features, the impact of different designs on the release of activity is indicated. The predictions of the proposed model are calculated for each stage and compared with the releases of activity predicted by the licensing methods that were used in the earlier benchmark studies. The conservative nature of the latter is confirmed. Methods and data are also presented for calculating the resulting doses to members of the public of the National Radiological Protection Boards as a result of work carried out by several national bodies in the UK. Other, equally acceptable, models are used in other countries of the Community and some examples are given

  16. Revisiting large break LOCA with the CATHARE-3 three-field model

    International Nuclear Information System (INIS)

    Valette, Michel; Pouvreau, Jerome; Bestion, Dominique; Emonot, Philippe

    2009-01-01

    Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in an isolated rod bundle mockup is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven Reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit : core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic 6-equation model is used in the other parts of the loop. A short analysis of the results is presented. (author)

  17. PELE-IC, 2-D Eulerian Incompressible Hydrodynamic and Bubble Dynamic after LWR LOCA

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.

    1981-01-01

    1 - Description of problem or function: PELE-IC is a two-dimensional semi-implicit Eulerian hydrodynamics program for the solution of incompressible flow coupled to flexible structures. The code was developed to calculate fluid-structure interactions and bubble dynamics of a pressure-suppression system following a loss-of- coolant accident (LOCA). The fluid, structure, and coupling algorithms have been verified by calculation of benchmark problems and air and steam blowdown experiments. The code is written for both plane and cylindrical coordinates. The coupling algorithm is general enough to handle a wide variety of structural shapes. The concepts of void fractions and interface orientation are used to track the movement of free surfaces, allowing great versatility in following fluid-gas interfaces both for bubble definition and water surface motion without the use of marker particles. 2 - Method of solution: The solution strategy is to first solve the Navier-Stokes equations explicitly using values from the previous time-step. Since these values do not necessarily satisfy the continuity equation, the pressure field is iterated upon until the incompressibility condition for each computational cell is satisfied within prescribed limits. The structural motion is computed by a finite element code from the applied pressure at the fluid-structure interface. The shell structure algorithm uses conventional thin-shell theory with transverse shear. The finite-element spatial discretization employs piecewise-linear interpolation functions and one-point quadrature applied to conical frustra. The Newmark implicit time integration method is used as a one-step module. The fluid code then uses the structure's position and velocity as boundary conditions. The fluid pressure field and the structure's response are corrected iteratively until the normal velocities of fluid and structure are equal. The effects of steam condensation and oscillatory chugging on structures are

  18. Evaluation on the habitability of a reactor control room for a 1300 MWe PWR following a LOCA

    International Nuclear Information System (INIS)

    Chang, Si Young; Ha, Chung Woo

    1988-01-01

    An evaluation on the habitability of a reactor control room for a French 1300 MWe P'4 type PWR following a LOCA has been performed through exposure dose assessment for a reactor operator. A computer code COREX calculating the time-integrated exposure dose has been developed to provide a reasonable basis in this evaluation. Using COREX the exposure dose reduction factors in the reactor control room, the time--integrated radioactivities released into the atmosphere and the time-integrated exposure dose up to 30 days following the LOCA can be also calculated. From the exposure dose assessment, the time-integrated exposure dose to whole body and thyroid of a reactor operator were 0.36 mSv(0.036 rem) and 480 mSv(48.0 rem), respectively after 30 days following the LOCA. The thyroid dose of 480 mSv was nearly 10 times greater than the dose equivalent limit of 50 mSv(5.0 rem) set by the ICRP. Regarding the habitability of a reactor control room, this exceeding thyroid exposure dose could be reduced to 1.2 mSv(0.12 rem), which is 400 times less than the original, by considering the practical 4 work-shifts a day, and by improving the iodine removal efficiency of the filtration system n the reactor control room through the reinforcement of charcoal bed filters for iodine removal. The radiological habitability of a reactor control room, therefore, could be assured by comparing with the dose equivalent limit of the ICRP

  19. Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

    International Nuclear Information System (INIS)

    Lelong, Franck

    2010-01-01

    In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

  20. Simulation of LOCA and ageing effect with containment liner mockup for analysis of liner-concrete interaction

    International Nuclear Information System (INIS)

    Wienand, B.; Fila, A.; Hermann, N.; Mueller, M.

    2015-01-01

    The investigation of the pre-stressed concrete wall behavior including the liner during LOCA conditions is important for the assessment of the structural integrity of the structure and the leak tightness of the liner. In the frame of the NUGENIA ACCEPPT project WP1 G4 'Structural interaction of liner with the concrete', a load test on a reactor containment liner mockup was carried out. The pre-stressed mockup represents a cylindrical part of the liner, embedded in the concrete wall, but without the wall curvature which is not test relevant. It correlates in material and geometrical properties to the EPR containment. The purpose of the test was to check the liners structural behavior and its integrity for Loss of Coolant Accident (LOCA) load combination considering pre-stressing forces and ageing effects due to creep and shrinkage including liner buckling. The test was carried out at the Karlsruhe Institute of Technology (KIT) in September 2013. This article presents the measurement technology, the results and the development of a calculation method for the embedded liner structure. It appears that the liner deformation results are exemplarily shown at the locations of the imperfections, where the liner buckling is anticipated. The measured liner surface strains ranged between +2 and -10 per thousand. The compressive strains are higher than the tensile strains due to the compressive membrane strains caused by pre-stressing and heating. Although the liner got plastic deformations, the liner strains are still far below the elongation at rupture, which indicates that the liner integrity is ensured. We can conclude that the liner mockup test proceeded as planned. The evaluation results show that the purpose of the liner mockup to simulate LOCA + ageing conditions and liner buckling has fully been achieved

  1. Analysis and development of the automated emergency algorithm to control primary to secondary LOCA for SUNPP safety upgrading

    International Nuclear Information System (INIS)

    Kim, V.; Kuznetsov, V.; Balakan, G.; Gromov, G.; Krushynsky, A.; Sholomitsky, S.; Lola, I.

    2007-01-01

    The paper presents the results of the study conducted to support planned modernization of the South Ukraine nuclear power plant. The objective of the analysis has been to develop the automated emergency control algorithm for primary to secondary LOCA accident for SUNPP WWER-1000 safety upgrading. According to the analyses performed in the framework of safety assesment report, given accident is the most complex for control and has the largest contribution into the core damage frequency value. This is because of initial event diagnostics is difficult, emergency control is complicated for personnel, time available for decision making and actions performing is limited with coolant inventory for make-up, probability of steam dump valves on affected steam generator non-closing after opening is high, and as a consequence containment bypass, irretrievable loss of coolant and radioactive materials release into the environment are possible. Unit design modifications are directed on expansion of safety systems capabilities to overcome given accident and to facilitate the personnel actions on emergency control. Safety systems modification according to developed algorithm will allow to simplify accident control by personnel and enable to control the ECCS discharge limiting pressure below the affected steam generator steam dump valve opening pressure, and decrease the probability of the containment bypass sequences. The analysis of the primary-to-secondary LOCA thermal-hydraulics has been conducted with RELAP5/Mod 3.2, and involved development of the dedicated analytical model, calculations of various plant response accident scenarios, conducting of plant personnel intervention analyses using full-scale simulator, development and justification of the emergency control algorithm aimed on the minimization of negative consequences of the primary-to-secondary LOCA (Authors)

  2. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  3. Strain measurements at the HDR-pipe-system under LOCA-load: Effects on elbows and displaced weldings

    International Nuclear Information System (INIS)

    Hunger, H.

    1985-01-01

    This paper characterizes some effects which have been detected during strain gauge measurements on a test piping with feed water check valve oscillating under blowdown-load. The ovalization of a pipe elbow subjected to in-plane-bending affects the connected straight pipe; this is shown by means of circumferential stresses. Very high LOCA-load produces plastic strain and changes the pipe dynamics. Artificial displaced welds increase the local strain but no defects have occurred. One example compares stresses from measurement and post-calculation. Moreover there are given some remarks on the optimization of the comparison of measurement and calculation. (orig.)

  4. The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

    International Nuclear Information System (INIS)

    Abdul-Razzak, A.; Zhang, J.; Sills, H.E.; Flatt, L.; Jenkins, D.; Wallace, D.J.; Popov, N.

    2002-01-01

    The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its proto-typing application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX R fuel. The methodology is consistent with and builds on world practice. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions. (authors)

  5. Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

    International Nuclear Information System (INIS)

    Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

    1988-12-01

    A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

  6. The Effect of External Vessel Cooling for a 2 inch LOCA Severe Accident Scenario at SMART with MIDAS/SMR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

    2010-01-01

    KAERI is developing a new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. This feature may prevent the large size of LOCA. However it is necessary to estimate the hypothetical severe accidents progression for improving the degree of safety and identifying the unknown weakness of the system against an accident. To simulate a hypothetical severe accident for the SMART, we adopt the MIDAS/SMR code which was developed by KAERI

  7. Frequency probabilistic analysis of a small break LOCA due to a power operated relief valve (PORV) for Angra-1 pre-TMI and post-TMI

    International Nuclear Information System (INIS)

    Onusic Junior, J.

    1986-01-01

    After the TMI event efforts were aimed towards improvements in the operational and administrative procedures related to the power operated relief valves (PORVs) in order to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by stuck-open power operated relief valve. This paper presents a frequency probabilistic analysis of a small break LOCA due to a stuck open PORV and safety valve to the Angra I nuclear power plant in operating conditions pre-TMI and post-TMI. (Author) [pt

  8. Comparison of computer codes (CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT) with data from the NRU-LOCA thermal hydraulic tests

    International Nuclear Information System (INIS)

    Mohr, C.L.; Rausch, W.N.; Hesson, G.M.

    1981-07-01

    The LOCA Simulation Program in the NRU reactor is the first set of experiments to provide data on the behavior of full-length, nuclear-heated PWR fuel bundles during the heatup, reflood, and quench phases of a loss-of-coolant accident (LOCA). This paper compares the temperature time histories of 4 experimental test cases with 4 computer codes: CE-THERM, FRAP-T5, GT3-FLECHT, and TRUMP-FLECHT. The preliminary comparisons between prediction and experiment show that the state-of-the art fuel codes have large uncertainties and are not necessarily conservative in predicting peak temperatures, turn around times, and bundle quench times

  9. THE PREDICTION OF pH BY GIBBS FREE ENERGY MINIMIZATION IN THE SUMP SOLUTION UNDER LOCA CONDITION OF PWR

    Directory of Open Access Journals (Sweden)

    HYOUNGJU YOON

    2013-02-01

    Full Text Available It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3&4 and UCN 1&2. As results, pH of the sump solution for the SKN 3&4 was between 7.02 and 7.45, and for the UCN 1&2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl, HNO3, and Cs are very low.

  10. Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    Rychkov, M.; Chikkanagoudar, U.; Sehgal, B.R.

    2004-01-01

    A RELAP5 model for the analysis of the PSB-VVER test facility was developed by EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, we have modified the PSB-VVER facility's RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5's calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the '11% UP LOCA' test data, the RELAP5/MOD3.2 model was used for a so-called 'blind' transient calculation of the test '2*25% HL LOCA' and the results obtained were compared with the experimental data provided after the calculation. From the results of the qualitative and quantitative comparison of the 2 test calculations and the experimental data, we can state that the RELAP5/MOD3.2 code gives a satisfactory modeling of the PSB-VVER facility' thermal hydraulic phenomena

  11. Assessment on the Reactor Containment Cooling Capability of Kori Unit 1 Under LOCA Conditions with Loss of Offsite Power

    International Nuclear Information System (INIS)

    Lee, Jin Yong; Park, Jong Woon; Kim, Hyeong Taek

    2006-01-01

    The fan cooler system is designed to remove heat from containment under postulated accident conditions. During a postulated LOCA concurrent with a Loss of Offsite Power (LOOP), the Component Cooling Water (CCW) pumps that supply cooling water to the fan cooler and the fan that supplies containment air to the fan cooler will temporarily lose power. Then, the high temperature steam in the containment atmosphere will pass over the fan cooler tubing without forced cooling water flow. In that case, boiling may occur in the fan cooler tubes causing steam bubbles to form and pass into the attached CCW piping creating steam voids. Prior to the CCW pumps restart, the presence of steam and subcooled water can induce the potential for water hammer. As the CCW pumps restart, the accumulated steam condenses and the pumped water can produce a water hammer when the void closes. The hydrodynamic loads caused by such a water hammer event could challenge the integrity and the function of the fan cooler and associated CCW system. With respect to this phenomena, the United States Nuclear Regulatory Commission (USNRC) issued the Generic Letter (GL) 96-06, which requests an assessment of the possibility of boiling and water hammer in the cooling water system. The objectives of this study are to develop a analysis method for predicting the thermal hydraulic status of containment fan cooler and then to assess the containment fan cooler of Kori Unit 1 using the developed model under a LOCA with LOOP

  12. Comparison of the DVI line break LOCA with the equivalent cold leg break with the ATLAS facility

    International Nuclear Information System (INIS)

    Choi, K. Y.; Cho, S.; Kang, K. H.; Park, H. S.; Kim, Y. S.; Baek, W. P.

    2010-01-01

    The APR1400 (Advanced Power Reactor, 1400 MWe) adopts a DVI (Direct Vessel Injection) method for ECC (Emergency Core Cooling) water delivery rather than a conventional CLI (Cold Leg Injection) method as an advanced safety feature. The break scenario of one DVI nozzle is taken into account in the small break LOCA analysis. Transient behavior during the DVI line breaks needs to be investigated and compared with the equivalent break on the cold leg. An 8.5-inch double-ended break of one DVI nozzle was simulated with the ATLAS, and a counterpart test for the DVI break was performed at the cold leg with the equivalent break size for comparison. This comparison will contribute to enhancing a comprehensive understanding of the thermal hydraulic behavior during transients. A constructed integral effect database is also used to validate the existing conservative safety analysis methodology and to develop a best-estimate safety analysis methodology for small-break LOCAs. A post-test calculation was performed with a best-estimate safety analysis code, MARS 3.1, in order to examine its prediction capability and to identify any code deficiencies for thermal hydraulic phenomena occurring during the transient. (authors)

  13. Best-estimated multi-dimensional calculation during LB LOCA for APR1400

    International Nuclear Information System (INIS)

    Oh, D. Y.; Bang, Y. S.; Cheong, A. J.; Woong, S.; Korea, W.

    2010-01-01

    Best-estimated (BE) calculation with uncertainty quantification for the emergency core cooling system (ECCS) performance analysis during Loss of Coolant Accident (LOCA) is more broadly used in nuclear industries and regulations. In Korea, demand on regulatory audit calculation is continuously increasing to support the safety review for life extension, power up-rating and advanced nuclear reactor design. The thermal-hydraulic system code, MARS (Multi-dimensional Analysis of Reactor Safety), with multi-dimensional capability is used for audit calculation. It achieves to describe the complicated phenomena in reactor coolant system by very effectively consolidating the one dimensional RELAP5/MOD3 with the multidimensional COBRA-TF codes. The advanced power reactors (APR1400) to be evaluated has four separated hydraulic trains of the high pressure injection system (HPSI) with direct vessel injection (DVI) which is different from the existing commercial PWRs. Also, the therma-hydraulic behavior of DVI plant would be considerably different from that of a cold-leg safety injection since the low pressure safety injection system are eliminated and the high pressure safety flow are injected into the specific elevation of reactor vessel downcomer. The ECCS bypass induced by the downcomer boiling due to hot wall heating of reactor vessel during reflooding phase is one of the important phenomena which should be considered in DVI plants. Therefore, in this study, BE calculation with one-dimensional (1-D) and multi-dimensional (multi-D) MARS models during LBLOCA are performed for APR1400 plant. In the multi-D evaluation, the reactor vessel is modeled by multi-D components and the specific treatment of flow path inside reactor vessel, e.g., upper guide structure, is essential. The concept of hot zone is adopted to simulate the limiting thermal-hydraulic conditions surrounding hot rod, which is similar to hot channel in 1-D. Also, alternative treatment of the hot rods in multi-D is

  14. Reconstruction of intra-bundle fission density profile during a postulated LOCA in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, D. [Oak Ridge National Laboratory (United States); Rahnema, F. [Georgia Inst. of Technology (United States); Nuclear and Radiological Engineering/Medical Physics Programs, George W. Woodruff School, Georgia Inst. of Technology, Atlanta, GA 30332-0405 (United States); Serghiuta, D. [Canadian Nuclear Safety Commission (Canada); Sarsour, H.; Turinsky, P. J. [North Carolina State Univ. (United States); Stamm' ler, R. [Studsvik Scandpower AS (Norway)

    2006-07-01

    In this paper, results related to the reconstruction of intra-bundle fission density profile for a 37-pin CANDU-6 bundle with the highest enthalpy deposition during a postulated large LOCA stagnation break in a Bruce B core are presented. Bruce B is a nuclear power plant in Kincardine, Ontario (Canada)), on the shores of Lake Huron with 4 CANDU reactors that are rated at about 750 MWe. The reconstruction of the fuel pin fission densities is based on steady-state, three-dimensional simulations with the Monte Carlo code MCNP for a subset of 27 out of 69 time steps during the first two seconds of the power pulse predicted for the fuel bundle at core location V13/8. Two-group cross section data libraries are generated for MCNP at each time step by the lattice depletion neutron transport code HELIOS-1.7. To include the effect of the surrounding core environment, the calculations are performed with time-dependent albedo boundary conditions inferred from a full core simulation of the transient by the nodal diffusion code NESTLE with HELIOS homogenized cross-sections. It is found that the local peaking factor (LPF) in the outer ring varies during the transient, but never exceeds its value before the transient. Inclusion of the core environment increases the LPF in the outer ring. For the analyzed case, the increase is 0.72% with a relative error of 0.01% for the LPF before the transient and 0.55% (with a relative error of 0.01%) for the maximum average LPF during the transient. The latter is based on only four selected transient time points. Note that the immediate environment of the 'hot bundle' does not contain any reactivity devices or other perturbing factors. As a result, the increases observed in the LPF in the outer ring may not be representative of the situations in which 'other' core environment perturbing factors are present. To determine the effect of these factors on the LPF, further analyses of a bundle in the proximity of control devices

  15. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A.; Elkin, I.V.; Pylev, S.S.

    2005-01-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  16. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A. [Elektrogorsk Research and Engineering Center, EREC, Bezymiannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Elkin, I.V.; Pylev, S.S. [NSI RRC ' Kurchatov Institute' , Kurchatov Sq., 1, Moscow, 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  17. Generating debate on issues surrounding the venting of containment in the event of LOCA to ensure an optimised safe outcome

    International Nuclear Information System (INIS)

    Chadwick, Chris; Jahouel, Xavier; Swain, Adam

    2014-01-01

    Following the incidents at the Japanese Fukushima Nuclear facility, when three units experienced LOCA, the consequences of those events have caused ripples across the world. Regulators around the world are examining the need to prevent excessive pressurisation of NPP Containment, and safely evacuate the gaseous consequences of LOCA, there is a need to return to fundamental principles and examine each putative set of data derived from the various models describing the consequence of LOCA and other events, a LOCA being a 'Loss Of Coolant Accident', and represents the outcome from a range of incidents ranging from fuel pellets being exposed to cooling (heat exchange) water, to a complete melt down of the fuel load with consequence evaporation of the concrete structures of the reactor containment buildings. Clearly it would be highly desirable in both economic and safety terms to minimise the external impact of these events inside the containment. Since one of the results of a LOCA is the pressurisation of the internal space of the containment building, regulators and the industry are looking at the mandatory installation of suitable vent systems to vent the pressure building up inside the containment, whilst ensuring the minimum impact on the surrounding environment. This is clearly a filtration/separation/recombination issue, and as an expert engineering company in the nuclear industry, Porvair has concerns that the issues of Containment Venting are not being addressed by expert filter companies, but by expert nuclear engineering companies with only a superficial knowledge of the complexities and nuances on filtration processes. The paper will describe in depth and detail the individual consequences of each particular aspect of the modelled data. Looking at flowing conditions (pressure, temperature, gas constituents including water vapour and Caesium and Iodine compounds), vent pressure philosophy, deposition of solids (size, type and quantity), decay heat

  18. International Standard Problems and Small Break Loss-Of-Coolant Accident (SBLOCA)

    International Nuclear Information System (INIS)

    Aksan, N.

    2008-01-01

    Best-estimate thermalhydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. In this respect, parallel to other national and international programs, OECD/Nea (OECD Nuclear Energy Agency) Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years some forty-eight International Standard Problems (ISPs). These ISPs were performed in different fields as in-vessel thermalhydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermalhydraulic behaviour. 80% of these ISPs were related to the working domain of Principal Working Group no. 2 on Coolant System Behaviour (PWG2). The ISPs have been one of the major PWG2 activities for many years. The individual ISP comparison reports include the analysis and conclusions of the specific ISP exercises. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISP's is given in this paper based on a report prepared by a CSNI-PWG2 writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident, specifically small break LOCA, are shortly summarized. Five small break LOCA related ISP's are considered, since these were used for the assessment of the advanced best-estimate codes. The considered ISP's deal with the phenomenon typical of small break LOCAs in Western design PWRs. The experiments in four integral test facilities, LOBI, SPES, BETHSY

  19. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  20. Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

    International Nuclear Information System (INIS)

    Khvostov, G.; Wiesenack, W.; Zimmermann, M.A.; Ledergerber, G.

    2011-01-01

    Highlights: → A model for the dynamics of axial gas redistribution in fuel rods during the LOCA is developed and coupled to the FALCON fuel behaviour code. → The first verification of the model is carried out using the data of the selected Halden LOCA tests. → According to calculation, the short rods used in the Halden tests show a small effect of the delayed gas redistribution during the clad ballooning. → The predicted effect is significant in the full length rods, eventually resulting in a considerable delay of the predicted moment of cladding rupture. → The predicted delay of cladding burst may be large enough to eventually affect the efficiency of the emergency core cooling system. - Abstract: A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code. The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from

  1. Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements

    International Nuclear Information System (INIS)

    Wilkowski, G.; Rudland, D.; Wolterman, R.; Krishnaswamy, P.; Scott, P.; Rahman, S.; Fairbanks, C.

    2002-01-01

    The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, A cceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors . These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from the

  2. Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs; Comparacion de la fenomenologia de las secuencias de SBO con y sin LOCA de sellos en reactores PWR Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

    2013-07-01

    SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

  3. Verification of human actions in SBO sequences with LOCA stamps in Westinghouse PWRs; Verificacion de las actuaciones humanas en secuencias de SBO con LOCA de sellos en reactores PWR Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Mena Rosell, L.; Jimenez Varas, G.

    2013-07-01

    The Fukushima accident has shown the need for tools and methodologies able to analyze human activities and / or capabilities of portable systems that has given the Spanish plants as a result of the stress tests . In this work we have applied the methodology of integrated safety analysis developed by the CSN , to SBO sequences with LOCA stamp. The aim is to show a methodology for testing the performances of the Emergency Operating Procedures and Guides Severe Accident Management. The simulations were performed with the tool SCAIS coupled to MAAP . The results show that there are human activities that may be beneficial in certain sequences but harmful in others. This type of problem is already known and referred to in the GGAS . However, FSR shows a practical way to check human actions cannot be obtained with other methods.

  4. Standard problem exercise to validate criticality codes for large arrays of packages of fissile materials

    International Nuclear Information System (INIS)

    Whitesides, G.E.; Stephens, M.E.

    1986-01-01

    A study has been conducted by an Office of Economic Cooperation and Development-Committee on the Safety of Nuclear Installations (OECD-CSNI) Working Group that examined computational methods used to compute k/sub eff/ for large greater than or equal to5 3 arrays of fissile material (in which each unit is a substantial fraction of a critical mass). Five fissile materials that might typically be transported were used in the study. The ''packages'' used for this exercise were simplified to allow studies unperturbed by the variety of structural materials which would exist in an actual package. The only material present other than the fissile material was a variation in the moderator (water) surrounding the fissile material. Consistent results were obtained from calculations using several computational methods. That is, when the bias demonstrated by each method for actual critical experiments was used to ''correct'' the results obtained for systems for which there were no experimental data, there was good agreement between the methods. Two major areas of concern were raised by this exercise. First, the lack of experimental data for arrays with size greater than 5 3 limits validation for large systems. Second, there is a distinct possibility that the comingling of two shipments of unlike units could result in a reduction of the safety margins. Additional experiments and calculations will be required to satisfactorily resolve the remaining questions regarding the safe transport of large arrays of fissile materials

  5. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall

  6. A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

    International Nuclear Information System (INIS)

    Ramani, D.T.

    1977-01-01

    The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

  7. Application of code scaling, applicability and uncertainty methodology to large break LOCA analysis of two loop PWR

    International Nuclear Information System (INIS)

    Mavko, B.; Stritar, A.; Prosek, A.

    1993-01-01

    In NED 119, No. 1 (May 1990) a series of six papers published by a Technical Program Group presented a new methodology for the safety evaluation of emergency core cooling systems in nuclear power plants. This paper describes the application of that new methodology to the LB LOCA analysis of the two loop Westinghouse power plant. Results of the original work were used wherever possible, so that the analysis was finished in less than one man year of work. Steam generator plugging level and safety injection flow rate were used as additional uncertainty parameters, which had not been used in the original work. The computer code RELAP5/MOD2 was used. Response surface was generated by the regression analysis and by the artificial neural network like Optimal Statistical Estimator method. Results were compared also to the analytical calculation. (orig.)

  8. Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code

    International Nuclear Information System (INIS)

    Nagler, A.; Gilat, J.; Hirshfeld, H.

    1991-01-01

    The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)

  9. A PC Mathcad-based computational aid for severe accident analysis and its application to a BWR small LOCA sequence

    International Nuclear Information System (INIS)

    Wu, Laung-Kuang T.; Lee, S.J.

    2004-01-01

    A PC-based Mathcad program is used to develop a computational aid for analyzing severe accident phenomena. This computational aid uses simple engineering expressions and empirical correlations to estimate key quantities and timings at various stages of accident progressions. In this paper, the computational aid is applied to analyze an early phase of a BWR small LOCA sequence. The accident phenomena analyzed include: break flow rates, boiled-up water level in the core, core uncovery time, depressurization of the reactor pressure vessel, core heat-up, onset of clad oxidation, hydrogen generation, and onset of fuel relocation. The results are compared with those obtained running the MAAP 3.0B code. This PC-based computational aid can be used to train plant personnel in understanding severe accident phenomena and to assist them in managing severe accidents. (author)

  10. Qualification of cables to IEEE standards 323-1974 and 383-1974

    International Nuclear Information System (INIS)

    Hosticka, C.; Kingsbury, E.R.; Bruhin, A.C.

    1980-01-01

    Wire and Cable manufacturers generally qualify products for class IE application by envelope type testing to user specifications and environmental conditions recommended by IEEE Standards 323-1974 and 383-1974. The General Electric Wire and Cable Business Department recently completed two such qualification programs. Cable constructions tested were 600V control cables and 600 V, 2kV, and 15kV power cables insulated with flame resistant mineral filled crosslinked polyethylene. The 15kV samples included taped field splices. In the second test program, the steam pressure-temperature profile included a simulated main steam line break. Test specimens were wrapped on grounded mandrels and were electrically loaded throughout the simulated LOCA tests. After completion of environmental testing, samples were subjected to the IEEE 383 simulated post-LOCA test. 6 refs

  11. Qualification of cables to IEEE standards 323-1974 and 383-1974

    International Nuclear Information System (INIS)

    Hosticka, C.; Kingsbury, E.R.; Bruhin, A.C.

    1980-01-01

    Wire and Cable manufacturers generally qualify products for class IE application by envelope type testing to user specifications and environmental conditions recommended by IEEE Standards 323-1974 and 383-1974. The General Electric Wire and Cable Business Department recently completed two such qualification programs. Cable constructions tested were 600V control cables and 600 V, 2KV, and 15KV power cables insulated with flame resistant mineral filled crosslinked polyethylene. The 15KV samples included taped field splices. In the second test program, the steam pressure-temperature profile included a simulated main steam line break. Test specimens were wrapped on grounded mandrels and were electrically loaded throughout the simulated LOCA tests. After completion of environmental testing, samples were subjected to the IEEE 383 simulated post-LOCA test. 6 refs

  12. Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

    International Nuclear Information System (INIS)

    Bujan, A.; Adamik, V.; Misak, J.

    1986-01-01

    A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

  13. Modeling the cool down of the primary heat transport system using shut down cooling system in normal operation and after events such as LOCA

    International Nuclear Information System (INIS)

    Icleanu, D.L.; Prisecaru, I.

    2015-01-01

    This paper aims at modeling the cooling of the primary heat transport system using shutdown cooling system (SDCS), for a CANDU 6 NPP in all operating modes, normal and abnormal (particularly in case of LOCA accident), using the Flowmaster calculation code. The modelling of heavy water flow through the shutdown cooling system and primary heat transport system was performed to determine the distribution of flows, pressure in various areas of the hydraulic circuit and the pressure loss corresponding to the components but also for the heat calculation of the heat exchangers related to the system. The results of the thermo-hydraulic analysis show that in all cases analyzed, normal operation and for LOCA accident regime, the performance requirements are confirmed by analysis

  14. Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of excedeence of damage by integrated Safety Analysis Methodology

    International Nuclear Information System (INIS)

    Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

    2013-01-01

    The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Excedence Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

  15. A thermohydraulic analysis for LOCA accident of a CANDU 600 reactor core charged with SEU 43 fuel by means of FIREBIRD code

    International Nuclear Information System (INIS)

    Serbanel, M.; Catana, A.

    2001-01-01

    This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters

  16. Appendix S-NH-1 and S-NH-2 of the experiment operating specification for the semiscale MOD-2C small break LOCA without HPI experiment series

    International Nuclear Information System (INIS)

    Owca, W.A.

    1985-10-01

    This document is Appendix S-NH--1 and S-NH-2 of the Experiment Operating Specification (EOS) for the Small Break LOCA without high pressure injection (HPI) series. It contains detailed information on the S-NH-1 and S-NH-2 experiment operation and facility configuration necessary to meet the series objectives stated in the main EOS body. 14 refs., 17 figs

  17. Verification of LOCA/ECCS analysis codes ALARM-B2 and THYDE-B1 by comparison with RELAP4/MOD6/U4/J3

    International Nuclear Information System (INIS)

    Shimizu, Takashi

    1982-08-01

    For a verification study of ALARM-B2 code and THYDE-B1 code which are the component of the JAERI code system for evaluation of BWR ECCS performance, calculations for typical small and large break LOCA in BWR were done, and compared with those by RELAP4/MOD6/U4/J3 code. This report describes the influences of differences between the analytical models incorporated in the individual code and the problems identified by this verification study. (author)

  18. UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

  19. ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R; Shen, K; Sjoeberg, A

    1995-03-01

    This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs.

  20. LOFA [loss of flow accident] and LOCA [loss of coolant accident] in the TIBER-II engineering test reactor: Appendix A-4

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510 0 C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs

  1. LOCA testing of high burnup PWR fuel in the HBWR. Additional PIE on the cladding of the segment 650-5

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Espeland, M.; Jenssen, H.K.

    2008-07-01

    IFA-650.5, a test with pre-irradiated fuel in the Halden Project LOCA test series, was conducted on October 23rd, 2006. The fuel rod had been used in a commercial PWR and had a high burnup, 83 MWd/kgU. Experimental arrangements of the fifth test were similar to the preceding LOCA tests. The peak cladding temperature (PCT) level was higher than in the third and fourth tests, 1050 C. A peak temperature close to the target was achieved and cladding burst occurred at approx. 750 C. Within the joint programme framework of the Halden Project PIE was done, consisting of gamma scanning, visual inspection, neutron-radiography, hydrogen analysis and metallography / ceramography. An additional extensive PIE including metallography, hydrogen analysis, and hardness measurements of cross-sections at seven axial elevations was done. It was completed to study the high burnup and LOCA induced effects on the Zr-4 cladding, namely the migration of oxygen into the cladding from the inside surface, the cladding distension, and the burst (author)(tk)

  2. Large Break LOCA Analysis with New downcomer Nodalizaion and Multi-Dimensional Model and Effect of Cross flow option in MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hyung-wook; Lee, Sang-yong; Oh, Seung-jong; Kim, Woong-bae [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    The phenomena of LOCA have been investigated for long time. The most extensive research project for LOCA was the 2D/3D program experiments. The results of the 2D/3D experiments show flow conditions in the downcomer during end-of-blowdown were highly multi-dimensional at full-scale. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. An LBLOCA analysis for APR1400 with new downcomer input deck was conducted using KREM with MARS-KS 1.4 Version code. Analysis was processed under LBCOCA of 100% break size of cold leg case. The authors developed input deck with new downcomer nodalizaion and Multi-Dimensional downcomer model, then implemented LOCA analysis with new input decks and compared with existing analysis results. PCT from new input and multi-dimensional input deck shows similar PCT trend from original input deck. There occurred more rapid drop of PCT from new and multidimensional input deck than original input deck. PCT from new and multidimensional input deck are satisfied with PCT design limit. It can be concluded that there occurs no acceptance criteria issue even though new and multidimensional input deck are applied to LBLOCA analysis. In future study, comparative analysis with experiment results will be implemented.

  3. ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

    International Nuclear Information System (INIS)

    Blomquist, R.; Shen, K.; Sjoeberg, A.

    1995-03-01

    This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs

  4. Preliminary accident analysis of Loss of Off-Site Power and In-Box LOCA for the CFETR helium cooled solid breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Lian, Qiang; Cui, Shijie [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Zhang, Jing; Zhang, Dalin; Su, G.H. [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China); Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi’an Jiaotong University, Xi’an, 710049 (China)

    2017-05-15

    Highlights: • The CFETR HCSB blanket has been investigated using RELAP5. • Loss of Off-Site Power is investigated. • The parametric analyses during In-Box LOCA are investigated. • The HCSB blanket for CFETR is designed with sufficient decay heat removal capability. - Abstract: As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage.

  5. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

    2013-10-01

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  6. An attempt for a unified description of mechanical testing on Zircaloy-4 cladding subjected to simulated LOCA transient

    Directory of Open Access Journals (Sweden)

    Desquines Jean

    2016-01-01

    Full Text Available During a Loss Of Coolant Accident (LOCA, an important safety requirement is that the reflooding of the core by the emergency core cooling system should not lead to a complete rupture of the fuel rods. Several types of mechanical tests are usually performed in the industry to determine the degree of cladding embrittlement, such as ring compression tests or four-point bending of rodlets. Many other tests can be found in the open literature. However, there is presently no real intrinsic understanding of the failure conditions in these tests which would allow translation of the results from one kind of mechanical testing to another. The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing. To achieve this objective, the failure mechanisms of pre-oxidized and pre-hydrided cladding samples are characterized by comparing the behavior of two different mechanical tests: Axial Tensile (AT test and “C”-shaped Ring Compression Test (CCT. The failure of samples in both cases can be described by usual linear elastic fracture mechanics theory. Using interrupted mechanical tests, metallographic examinations have evidenced that a set of parallel cracks are nucleated at the inner and outer surface of the samples just before failure, crossing both the oxide layer and the oxygen rich alpha layer. The stress intensity factors for multiple crack geometry are determined for both AT and CCT samples using finite element calculations. After each mechanical test performed on high temperature steam oxidized samples, metallography is then used to individually determine the crack depth and crack spacing. Using these two important parameters and considering the applied load at fracture, the stress intensity factor at failure is derived for each tested

  7. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

    Directory of Open Access Journals (Sweden)

    Li Yuquan

    2017-02-01

    Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

  8. Comparative experiments to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Li, YuQuan; Hao, Botao; Zhong, Jia; Wan Nam [State Nuclear Power Technology R and D Center, South Park, Beijing Future Science and Technology City, Beijing (China)

    2017-02-15

    The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME)—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative

  9. Application of realistic (best- estimate) methodologies for large break loss of coolant (LOCA) safety analysis: licensing of Westinghouse ASTRUM evaluation model in Spain

    International Nuclear Information System (INIS)

    Lage, Carlos; Frepoli, Cesare

    2010-01-01

    When the LOCA Final Acceptance Criteria for Light Water Reactors was issued in Appendix K of 10CFR50 both the USNRC and the industry recognized that the rule was highly conservative. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the USNRC began a research program to identify the degree of conservatism in those models permitted in the Appendix K rule and to develop improved thermal-hydraulic computer codes so that realistic accident analysis calculations could be performed. The overall results of this research program quantified the conservatism in the Appendix K rule and confirmed that some relaxation of the rule can be made without a loss in safety to the public. Also, from a risk-informed perspective it is recognized that conservatism is not always a complete defense for lack of sophistication in models. In 1988, as a result of the improved understanding of LOCA phenomena, the USNRC staff amended the requirements of 10 CFR 50.46 and Appendix K, 'ECCS Evaluation Models', so that a realistic evaluation model may be used to analyze the performance of the ECCS during a hypothetical LOCA. Under the amended rules, best-estimate plus uncertainty (BEPU) thermal-hydraulic analysis may be used in place of the overly prescriptive set of models mandated by Appendix K rule. Further guidance for the use of best-estimate codes was provided in Regulatory Guide 1.157 To demonstrate use of the revised ECCS rule, the USNRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology as an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis. More recently the CSAU principles have been generalized in the Evaluation Model Development and Assessment Process (EMDAP) of Regulatory Guide 1.203. ASTRUM is the Westinghouse Best Estimate Large Break LOCA evaluation model applicable to two-, three

  10. Development of gamma-ray densitometer and measurement of void fraction in instantaneous pipe rupture under BWR LOCA condition

    International Nuclear Information System (INIS)

    Yano, Toshikazu

    1983-11-01

    In order to clarify the transient mass flow rate under the instantaneous pipe rupture condition, it is necessary to use a highly sensitive void meter. Therefore, a high-response gamma-ray densitometer was developed for the measurement of void fraction variation caused by flashing vaporization of the high-pressure and -temperature water under the instantaneous pipe rupture accident. The measurement of void fraction was performed in the pipe rupture test under the BWR LOCA condition with a 6-inch diameter pipe. Initial conditions of the water were 6.86 MPa in pressure and the saturation temperature. To prove the reliability and accuracy, a calibration test by falling acrylic void simulators and an air injection test into cold water filled in the pipe were also conducted. The following results are obtained in the pipe rupture test. (1) The cone slit method is very useful to increase the measuring accuracy. (2) It is clearly observed that the apparent increase of void fraction occurs after the rarefaction wave passes. (3) The first maximum of void fraction occurs with some delay time after break. The following minimum void fraction concurs with the maximum pressure in the pressure recovering phenomena and with the maximum blowdown thrust force. (author)

  11. TRACG prediction of gravity-driven cooling system response in the SBWR/GIST facility LOCA tests

    International Nuclear Information System (INIS)

    Alamgir, M.; Andersen, J.G.M.; Yang, A.I.; Shiralkar, B.S.

    1990-01-01

    General Electric (BE) Nuclear Energy has initiated work on technology programs in support of the advanced light water reactor (ALWR) plants under contract to the U.S. Department of Energy (DOE). Work has been performed under the advanced boiling water reactor (ABWT) design verification program and the simplified boiling water reactor (SBWR) program. The objective of the SBWR program is to develop the key features of a simplified reactor design. The gravity-driven cooling system (GDCS) is an important feature of the SBWR design. The main objectives of the GDCS test program at GE were to demonstrate the technical feasibility of the GDCS concept by performing a section-scaled integrated systems test of the SBWR design and to provide a data base to qualify the TRACG computer code for use in SBWR accident analysis. This paper describes the qualification of TRACG for GDCS applications. The calculational capability and analytical models of TRACG are tested by performing assessment analysis for five loss-of-coolant-accident (LOCA) tests in the GDCS Integrated Systems Test (GIST) facility. The results of the qualification comparisons are presented and TRACG application ranges are discussed

  12. ICECON: a computer program used to calculate containment back pressure for LOCA analysis (including ice condenser plants)

    International Nuclear Information System (INIS)

    1976-07-01

    The ICECON computer code provides a method for conservatively calculating the long term back pressure transient in the containment resulting from a hypothetical Loss-of-Coolant Accident (LOCA) for PWR plants including ice condenser containment systems. The ICECON computer code was developed from the CONTEMPT/LT-022 code. A brief discussion of the salient features of a typical ice condenser containment is presented. Details of the ice condenser models are explained. The corrections and improvements made to CONTEMPT/LT-022 are included. The organization of the code, including the calculational procedure, is outlined. The user's manual, to be used in conjunction with the CONTEMPT/LT-022 user's manual, a sample problem, a time-step study (solution convergence) and a comparison of ICECON results with the results of the NSSS vendor are presented. In general, containment pressure calculated with the ICECON code agree with those calculated by the NSSS vendor using the same mass and energy release rates to the containment

  13. Safety analysis on Non-LOCA events for the revision of Wolsong NPP unit 2,3,4 sar

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Ryu, Eui Seung; Kho, Dong Wook; Kim, Sung Min

    2015-01-01

    Korean Wolsong Nuclear Power Plant Units 2,3,4 (CANDU-6 Type) has prepared the revision of safety analysis report (Final Safety Analysis Report (FSAR) chapter 15) from the original performed in the year of 1990s, using the updated and state-of-the-art methodology and tools including IST safety analysis codes and more detail modelling. Compared with the original FSAR15, the revised FSAR15 has significant improvement in both the scope and the depth of safety analysis, which has demonstrated the safety analysis results have complied with the safety requirements(acceptance criteria). This paper will present the analysis scope for Non-LOCA events re-analyzed or added for the FSAR15 revision, methodologies applied such as codes and modelling and some important analysis results will be demonstrated with comparison to acceptance criteria. Application of more detail and near-realistic assumptions and method including Dev-PDO options and uncertainty related to the CHF correlations has altogether brought about more safety margin compared with the original FSAR15 with respect to SDS trip effectiveness etc. (author)

  14. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  15. A Study of Time Response for Safety-Related Operator Actions in Non-LOCA Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Seok; Lee, Sang Seob; Park, Min Soo; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E and C Company, Daejeon (Korea, Republic of)

    2014-10-15

    The classification of initiating events for safety analysis report (SAR) chapter 15 is categorized into moderate frequency events (MF), infrequent events (IF), and limiting faults (LF) depending on the frequency of its occurrence. For the non-LOCA safety analysis with the purpose to get construction or operation license, however, it is assumed that the operator response action to mitigate the events starts at 30 minutes after the initiation of the transient regardless of the event categorization. Such an assumption of corresponding operator response time may have over conservatism with the MF and IF events and results in a decrease in the safety margin compared to its acceptance criteria. In this paper, the plant conditions (PC) are categorized with the definitions in SAR 15 and ANS 51.1. Then, the consequence of response for safety-related operator action time is determined based on the PC in ANSI 58.8. The operator response time for safety analysis regarding PC are reviewed and suggested. The clarifying alarm response procedure would be required for the guideline to reduce the operator response time when the alarms indicate the occurrence of the transient.

  16. Assessment of current safety evaluation analysis on reflood behavior during PWR-LOCA by using CCTF data

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Murao, Yoshio; Akimoto, Hajime; Okubo, Tsutomu; Sugimoto, Jun; Hojo, Tsuneyuki.

    1987-01-01

    This paper describes an assessment result on conservatism of current safety analysis concerning reflood behavior during a LOCA in a PWR by using the experimental data with cylindrical core test facility (CCTF) performed at Japan Atomic Energy Research Institute (JAERI). WREM code is selected for a representative of current safety analyses. The predicted peak clad temperature with the WREM code was higher than the data, and it was confirmed that the WREM code had the overall conservatism against CCTF data. The WREM code predicted the reasonable core boundary conditions and it was found that the conservatism of the code came mainly from the calculations on the incore thermal hydraulics and clad temperature. In addition, it was found that the conservatism of the WREM code against the CCTF data could be attributed to the neglection of horizontal fluid mixing between subchannels, the neglection of the heat transfer enhancement due to the radial core power profile, and the usage of the heat transfer correlations conservative against CCTF data. (author)

  17. Temporary core liquid level depression during cold-leg small-break LOCA effect of break size and power level

    International Nuclear Information System (INIS)

    Koizumi, Y.; Kumamaru, H.; Mimura, Y.; Kukita, Y.; Tasaka, K.

    1989-01-01

    Cold-leg small break LOCA experiments (0.5-10% break) were conducted at the large scale test facility (LSTF), a volumetrically-scaled (1/48) simulator of a PWR, of the ROSA-IV Program. When a break area was less than 2.5% of the scaled cold-leg flow area, the core liquid level was temporarily further depressed to the bottom elevation of the crossover leg during the loop seal clearing early in the transient only by the manometric pressure balance since no coolant remained in the upper portion of the primary system. When the break size was larger than 5%, the core liquid level was temporarily further depressed lower than the bottom elevation of the crossover leg during the loop seal clearing since coolant remained at the upper portion of the primary system; the steam generator (SG) U-tube upflow side and the SG inlet plenum, due to counter current flow limiting by updrafting steam while the coolant drained. The amount of coolant trapped there was dependent on the vapor velocity (core power); the larger the core power, the lower the minimum core liquid level. The RELAP5/MOD2 code reasonable predicted phenomena observed in the experiments. (orig./DG)

  18. Corrosion of aluminum and zinc in containment following a LOCA and potential for precipitation of corrosion products in the sump

    International Nuclear Information System (INIS)

    Niyogi, K.K.; Lunt, R.R.; Mackenzie, J.S.

    1982-01-01

    Following a loss-of-coolant accident (LOCA) in a LWR containment, certain materials in the containment come in contact with alkaline emergency cooling and containment spray solutions and may corrode yielding hydrogen gas. The problems associated with the production of hydrogen gas and the control of combustible gas concentration have been extensively explored in recent years. However, the phenomenon of corrosion and its consequences in the long term cooling of the reactor and the containment have drawn very little attention. United Engineers and Constructors Inc. has made an extensive effort to study through literature survey the solubility of the corrosion products from aluminum and zinc in order to assess the potential for precipitation in the containment sump. The analysis presented in this article is based on parameters for a typical large dry reactor containment with caustic/boric acid buffered spray solution. Parameters used in this study may vary from one plant to another. However, they are not expected to affect the overall conclusions

  19. Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

    Energy Technology Data Exchange (ETDEWEB)

    Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

    2006-07-01

    A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

  20. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    International Nuclear Information System (INIS)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6'' cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author)

  1. Cooling the intact loop of primary heat transport system using shut down cooling system after events such as LOCA

    International Nuclear Information System (INIS)

    Icleanu, D.L.

    2015-01-01

    The purpose of this paper is to model the Shutdown Cooling System operation for CANDU 6 NPP in case of LOCA accident, using Flowmaster calculation code by delimiting models and setting calculation assumptions and input data for hydraulic analysis, and and assumptions for the calculation and input data for calculating thermal performance check heat exchangers that are part of this system. The Flowmaster V7.8 code provides system engineers with a powerful tool to investigate pressure surge, pressure drop, flow rate, temperature and system response times - removing the uncertainty from fluid flow systems. Flowmaster is a one-dimensional thermal-hydraulic calculation code for dimensioning, analyzing and verifying the pipeline systems operation. Each component of Flowmaster is a mathematical model for an equipment that is included in a facility. Selected components are connected via nodes in order to form a network, which constitutes a computerized model of the system. Analyzing the parameters of the cooling system for all cooling processes considered it was found that the values obtained for thermal-hydraulic parameters, as well as the duration up to reaching specified limits fall within the design values of the system. This document is made up of an abstract and the slides of the presentation

  2. A Review of Dangerous Dust in Fusion Reactors: from Its Creation to Its Resuspension in Case of LOCA and LOVA

    Directory of Open Access Journals (Sweden)

    Andrea Malizia

    2016-07-01

    Full Text Available The choice of materials for the future nuclear fusion reactors is a crucial issue. In the fusion reactors, the combination of very high temperatures, high radiation levels, intense production of transmuting elements and high thermomechanical loads requires very high-performance materials. Erosion of PFCs (Plasma Facing Components determines their lifetime and generates a source of impurities (i.e., in-vessel tritium and dust inventories, which cool down and dilute the plasma. The resuspension of dust could be a consequences of LOss of Coolant Accidents (LOCA and LOss of Vacuum Accidents (LOVA and it can be dangerous because of dust radioactivity, toxicity, and capable of causing an explosion. These characteristics can jeopardize the plant safety and pose a serious threat to the operators. The purpose of this work is to determine the experimental and numerical steeps to develop a numerical model to predict the dust resuspension consequences in case of accidents through a comparison between the experimental results taken from campaigns carried out with STARDUST-U and the numerical simulation developed with CFD codes. The authors in this work will analyze the candidate materials for the future nuclear plants and the consequences of the resuspension of its dust in case of accidents through the experience with STARDUST-U.

  3. Dynamic event Tress applied to sequences Full Spectrum LOCA. Calculating the frequency of exceedance of damage by integrated Safety Analysis Methodology; Arboles de sucesos dinamicos aplicados a secuencias Full Spectrum LOCA. Calculo de la frequencia de excedencia del dano mediante la metodologia Analisis Integrados de Seguridad (ISA)

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Magan, J. J.; Fernandez, I.; Gil, J.; Marrao, H.; Queral, C.; Gonzalez-Cadelo, J.; Montero-Mayorga, J.; Rivas, J.; Ibane-Llano, C.; Izquierdo, J. M.; Sanchez-Perea, M.; Melendez, E.; Hortal, J.

    2013-09-01

    The Integrated Safety Analysis (ISA) methodology, developed by the Spanish Nuclear Safety council (CSN), has been applied to obtain the dynamic Event Trees (DETs) for full spectrum Loss of Coolant Accidents (LOCAs) of a Westinghouse 3-loop PWR plant. The purpose of this ISA application is to obtain the Damage Exceedance Frequency (DEF) for the LOCA Event Tree by taking into account the uncertainties in the break area and the operator actuation time needed to cool down and de pressurize reactor coolant system by means of steam generator. Simulations are performed with SCAIS, a software tool which includes a dynamic coupling with MAAP thermal hydraulic code. The results show the capability of the ISA methodology to obtain the DEF taking into account the time uncertainty in human actions. (Author)

  4. Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage.

  5. About the use of the CATHARE code for best estimate Large Break LOCA calculations: benefits for Safety and constraints

    International Nuclear Information System (INIS)

    Vacher, J.L.

    1994-01-01

    Since 1979, EDF has participated to the development of CATHARE, a best estimate accidental thermalhydraulic code, in collaboration with CEA and FRAMATOME. EDF is now investigating the use of this code for licensing studies and particularly for Large Break LOCA calculations. Until now, the work done at EDF, in relation with FRAMATOME and CEA, has mainly focused on the physical analysis of the transient and on the identification of the key phenomena. This task is a necessary step before uncertainty evaluation. To illustrate this point of view, a peculiar example of calculated Large Break transients for a 900 MW three loop plant is presented. In one of these calculations, a high value of Peak Cladding Temperature was obtained. This peculiar scenario was initiated by a large entrainment of water to the steam generators at the very beginning of the reflooding stage, followed by a strong pressurization which led to a lasting draining of the reactor vessel. The physical phenomena which determine the existence and amplitude of this scenario were identified and their influence was explained: condensation at the accumulator injection, heat exchange in the core, entrainment process to the steam generators. It appeared obvious that the large observed uncertainty was associated to only a few parameters. Although this peculiar system behaviour was obtained for only a particular combination of parameters and a narrow range of thermalhydraulic conditions, the capability of the code to simulate these phenomena was investigated in regard to experimental data. It was concluded that this scenario was definitely unrealistic on a reactor. Nevertheless, this peculiar example tends to demonstrate, firstly, that the use of a best-estimate code improves Safety as it makes possible to point out physical phenomena that could not be considered when using non mechanistic codes, secondly, that the uncertainty evaluation must be guided by a pertinent physical analysis of the transient, focusing

  6. Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

    International Nuclear Information System (INIS)

    Bang, Young Seok

    2015-01-01

    The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage

  7. Transient simulation of feedwater vaporization during a DBA LOP/LOCA using RELAP5/MOD3.1

    International Nuclear Information System (INIS)

    Harrell, J.R.; Fuller, R.W.

    1996-01-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station (GGNS) are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. The original design and testing requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. Given this condition, the appropriate testing criteria would be based on air with a relatively tight allowable limit. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leakage flow exists from the reactor vessel to the condenser through the feedwater piping during the reactor vessel blowdown phase. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

  8. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde; Simulacion de escenarios de accidente severo tipo perdida de refrigerante (Loca) pequeno, con el codigo MELCOR version 2.1 para la central nucleo-electrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V., E-mail: Jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft{sup 2} (4.6 cm{sup 2}), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

  9. Preliminary results of the BTF-104 experiment: an in-reactor test of fuel behaviour and fission-product release and transport under LOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, L W; Elder, P H; Devaal, J W; Irish, J D; Yamazaki, A R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    The BTF-104 experiment is one of a series of in-reactor tests being performed to measure fuel behaviour and fission-product release from nuclear fuel subjected to accident conditions. The primary objective of the BTF-104 experiment was to measure fission-product releases from a CANDU-sized fuel element under combined Loss-of-Coolant Accident (LOCA) and Loss-of-Emergency-Core-Cooling (LOECC) conditions at an average fuel temperature of about 1550 deg C. The preliminary results of the BTF-104 experiment are presented in this paper. (author). 6 refs., 12 figs.

  10. The NEPTUN experiments on LOCA thermal-hydraulics for tight-lattice PWRs

    International Nuclear Information System (INIS)

    Dreier, J.; Chawla, R.; Rouge, N.; Yanar, S.

    1990-01-01

    The NEPTUN test facility at the Paul Scherrer Institute is currently being used to provide a broad data base for the validation of thermal-hydraulics codes used in predicting the reflooding behaviour of a tight-lattice PWR (light water highb conversion reactor, LWHCR). The present paper gives a description of the facility and the matrix to be covered in the experimental program. Results are presented from a number of forced-feed, bottom-reflooding experiments, comparisons being made with (a) measurements carried out earlier for standard-PWR geometry and (b) the results of a calculational benchmark exercise conducted in the framework of a Swiss/German LWHCR-development agreement. Rewetting for the tight, hexagonal-geometry (p/d = 1.13) NEPTUN-III test bundle has been found to occur in all tests carried out to date, in which reasonably LWHCR-representative values for the various thermal-hydraulics parameters are used. Results of the calculational benchmark exercise have confirmed the need for further code development efforts for achieving reliable predictions of LWHCR reflooding behaviour. (author) 11 figs., 3 tabs., 3 refs

  11. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  12. Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

    International Nuclear Information System (INIS)

    Kolev, N.I.; Sybotinov, L.S.

    1984-01-01

    The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

  13. Evaluation of the radiative transfer in the core of a Pressurized Water Reactor (PWR) during the reflooding step of a Loss Of Coolant Accident (LOCA)

    International Nuclear Information System (INIS)

    Gerardin, J.

    2012-01-01

    We developed a method of resolution of radiative transfer inside a medium of vapor-droplets surrounded by hot walls, in order to couple it with a simulation of the flow at the CFD scale. The scope is the study of the cooling of the core of nuclear reactor following a Loss Of Coolant Accident (LOCA). The problem of radiative transfer can be cut into two sub problems, one concerning the evaluation of the radiative properties of the medium and a second concerning the solution of the radiative transfer equation. The radiative properties of the droplets have been computed with the use of the Mie Theory and those of the vapor have been computed with a Ck model. The medium made of vapor and droplets is an absorbing, anisotropically scattering, emissive, non grey, non homogeneous medium. Hence, owing to the possible variations of the flow properties (diameter and volumetric fraction of the droplets, temperature and pressure of the vapor), the medium can be optically thin or thick. Consequently, a method is required which solves the radiative transfer accurately, with a moderate calculation time for all of these prerequisites. The IDA has been chosen, derived from the well-known P1-approximation. Its accuracy has been checked on academical cases found in the literature and by comparison with experimental data. Simulations of LOCA flows have been conducted taking account of the radiative transfer, evaluating the radiative fluxes and showing that radiative transfer influence cannot be neglected. (author)

  14. Evaluation of PWR response to main-steamline break with concurrent steam-generator tube rupture and small-break LOCA

    International Nuclear Information System (INIS)

    Laaksonen, J.T.; Sheron, B.W.

    1982-12-01

    In 1980, the NRC staff raised a potential safety issue involving a coincident steamline break, steam generator tube rupture, and small-break loss-of-coolant accident (LOCA). The bases for this concern were that the system response, primarily the maintenance of core cooling, was unanalyzed and the adequacy of the present guidance to operators to respond to combination LOCAs was unknown. This report discusses the staff evaluations performed to assess the system response and the adequacy of the present emergency operator guidelines. In all of the analyzed cases the primary coolant shrinkage, caused by overcooling, and the simultaneous loss of coolant can be compensated by the high pressure emergency core cooling system. The core remains covered with liquid, and the primary coolant remains subcooled, except in the vessel upper head. If the steamline break is outside the containment and cannot be isolated, the radiological consequences could be more severe than in any accident currently analyzed in a typical plant Final Safety Evaluation Report (FSAR). To decrease the risk of elevated offsite releases, an early diagnosis of the tube rupture has to be ensured. This can be done by upgrading operator instructions. The appropriate mitigating actions are in the existing instructions

  15. Influence of partial blockage of a BWR bundle on heat transfer, cladding temperature, and quenching during bottom flooding or top spraying under simulated LOCA conditions

    International Nuclear Information System (INIS)

    Brand, B.; Gaul, H.P.; Sarkar, J.

    1982-01-01

    In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m 3 /h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800 0 C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis

  16. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Bolt, S.E.

    1977-01-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions

  17. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

    2008-01-01

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  18. Church ladies, good girls, and locas: stigma and the intersection of gender, ethnicity, mental illness, and sexuality in relation to HIV risk.

    Science.gov (United States)

    Collins, Pamela Y; von Unger, Hella; Armbrister, Adria

    2008-08-01

    Inner city women with severe mental illness may carry multiple stigmatized statuses. In some contexts these include having a mental illness, being a member of an ethnic minority group, being an immigrant, being poor, and being a woman who does not live up to gendered expectations. These potentially stigmatizing identities influence both the way women's sexuality is viewed and their risk for HIV infection. This qualitative study applies the concept of intersectionality to facilitate understanding of how these multiple identities intersect to influence women's sexuality and HIV risk. We report the firsthand accounts of 24 Latina women living with severe mental illness in New York City. In examining the interlocking domains of these women's sexual lives, we find that the women seek identities that define them in opposition to the stigmatizing label of "loca" (Spanish for crazy) and bestow respect and dignity. These identities have unfolded through the additional themes of "good girls" and "church ladies". Therefore, in spite of their association with the "loca", the women also identify with faith and religion ("church ladies") and uphold more traditional gender norms ("good girls") that are often undermined by the realities of life with a severe mental illness and the stigma attached to it. However, the participants fall short of their gender ideals and engage in sexual relationships that they experience as disempowering and unsatisfying. The effects of their multiple identities as poor Latina women living with severe mental illness in an urban ethnic minority community are not always additive, but the interlocking effects can facilitate increased HIV risks. Interventions should acknowledge women's multiple layers of vulnerability, both individual and structural, and stress women's empowerment in and beyond the sexual realm.

  19. Influence of chemical composition of zirconium alloy E110 on embrittlement under LOCA conditions - Part 1: Oxidation kinetics and macrocharacteristics of structure and fracture

    Science.gov (United States)

    Nikulin, S. A.; Rozhnov, A. B.; Belov, V. A.; Li, E. V.; Glazkina, V. S.

    2011-11-01

    Exploratory investigations of the influence of alloying and impurity content in the E110 alloy cladding tubes on the behavior under conditions of Loss of Coolant Accidents (LOCA) has been performed. Three alloys of E110 type have been tested: E110 alloy of nominal composition Zr-1%Nb (E110), E110 alloy of modified composition Zr-1%Nb-0.12%Fe-0.13%O (E110M), E110 alloy of nominal composition Zr-1%Nb with reduced impurity content (E110G). Alloys E110 and E110M were manufactured on the electrolytic basis and alloy E110G was manufactured on the basis of zirconium sponge. The high temperature oxidation tests in steam ( T = 1100 °C, 18% of equivalent cladding reacted (ECR)) have been conducted, kinetics of oxidation was investigated. Quantitative research of structure and fracture macrocharacteristics was performed by means of optical and electron microscopy. The results received were compared with the residual ductility of specimens. The results of the investigation showed the existence of "breakaway oxidation" kinetics and white spalling oxide in E110 and E110M alloys while the specimen oxidation kinetics in E110G alloy was characterized by a parabolic law and specimens had a dense black oxide. Oxygen and iron alloying in the E110 alloy positively changed the macrocharacteristics of structure and fracture. However, in general, it did not improve the resistance to embrittlement in LOCA conditions apparently because of a strong impurity influence caused by electrolytic process of zirconium production.

  20. Contribution from twenty two years of CSNI International Standard Problems

    International Nuclear Information System (INIS)

    1998-03-01

    This report provides a brief overview on the contribution of some CSNI International Standard Problems (ISPs) to nuclear reactor safety issues (41 ISPs performed over the last 22 years). This CSNI activity on ISPs has been one of the major activities of the Principal Working Group no.2 on Coolant System Behaviour. Its domain extended from thermal-hydraulics to several other accident domains following the main concerns of nuclear reactor safety, e.g., LOCA predictions fuel behaviour, operator procedures, containment thermal-hydraulics severe accidents, VVERs, etc. ISPs are providing unique material and benefits for some safety related issues. Clearly, all the technical findings and benefits provided by ISPs are still needed and contribute to advancement of nuclear safety. The report provides some overview on the general objectives of ISPs, content and types of ISPs, and technical domains covered by ISPs, followed by a synthesis of technical findings and benefits to the scientific community

  1. ['Gold standard', not 'golden standard'

    NARCIS (Netherlands)

    Claassen, J.A.H.R.

    2005-01-01

    In medical literature, both 'gold standard' and 'golden standard' are employed to describe a reference test used for comparison with a novel method. The term 'gold standard' in its current sense in medical research was coined by Rudd in 1979, in reference to the monetary gold standard. In the same

  2. Influence of the heat transfer to the building structures on the confinement parameters in case of a 200 mm LOCA on units 3 and 4 of NPP Kozloduy with reactors WWER-440/V230

    International Nuclear Information System (INIS)

    Stanev, I.

    2001-01-01

    The authors present a general analysis of the mass and energy balance in the WWER-440/V230 confinement structure in case of a 200 mm LOCA, based on results from CONTAIN calculations. The character of the heat transfer to the confinement building structures is analyzed and its influence on the confinement parameters is evaluated. The results show that the building structures heat consumption is comparable to that of the Spray System heat exchanger. (author)

  3. Decommissioning standards

    International Nuclear Information System (INIS)

    Crofford, W.N.

    1980-01-01

    EPA has agreed to establish a series of environmental standards for the safe disposal of radioactive waste through participation in the Interagency Review Group on Nuclear Waste Management (IRG). One of the standards required under the IRG is the standard for decommissioning of radioactive contaminated sites, facilities, and materials. This standard is to be proposed by December 1980 and promulgated by December 1981. Several considerations are important in establishing these standards. This study includes discussions of some of these considerations and attempts to evaluate their relative importance. Items covered include: the form of the standards, timing for decommissioning, occupational radiation protection, costs and financial provisions. 4 refs

  4. DRACCAR, a new 3D-thermal mechanical computer code to simulate LOCA transient on nuclear power plants. Status of the development and the validation

    International Nuclear Information System (INIS)

    Georges, Repetto; Francois, Jacq; Francois, Barre; Francois, Lamare; Jean-Marc, Ricaud

    2009-01-01

    IRSN is developing the DRACCAR computational software within the scope of its safety analyses on pressurised water reactors (PWR). This software is used to study loss-of-coolant accidents in the reactor core (LOCA) or in a spent fuel storage tank, for example. During such an accident, the coolant vaporises and the fuel rods dry out, which leads to an increase of their temperature, a swelling and fuel cladding failure. This swelling is responsible for major blockage in port of the core and can jeopardize the possibility of core cooling by means of back-up systems. The 3D multi-rod software is designed to model a fuel assembly so as to assess rod cooling and the blockage rate caused by deformed rods, by taking into account mechanical and thermal interactions between rods. The software can provide a consistent interpretation of the entire experimental database for a 'single-rod' configuration or a 'rod-bundle' configuration with either real or simulator fuel, transpose these results onto a reactor scale to determine what kind of research still needs to be conducted and finally, carry out safety studies. The models developed for this software cover: Heat transfers by conduction, convection and radiation. Oxidation of Zircaloy elements (cladding, guide tubes, inner shroud layer..) as well as hydriding process which can change mechanical properties. Thermomechanical behavior of fuel cladding (deformation and failure), including bowing phenomenon. Thermohydraulics on the scale of an assembly (to couple with an appropriate software), including a reflooding model. Fuel relocation and release of fission gases. A first version (DRACCAR V1) was delivered in March 2008 and is being validated on the basis of available experimental data (EDGAR, PHEBUS LOCA, PERICLES, REBEKA, HALDEN, etc.). A second version will be released in 2012 for which a coupling, in particular in the frame of the European NURISP project, is planned to an advanced sub-channel thermal-hydraulics code CATHARE

  5. Accounting standards

    NARCIS (Netherlands)

    Stellinga, B.; Mügge, D.

    2014-01-01

    The European and global regulation of accounting standards have witnessed remarkable changes over the past twenty years. In the early 1990s, EU accounting practices were fragmented along national lines and US accounting standards were the de facto global standards. Since 2005, all EU listed

  6. Standardization Documents

    Science.gov (United States)

    2011-08-01

    Specifications and Standards; Guide Specifications; CIDs; and NGSs . Learn. Perform. Succeed. STANDARDIZATION DOCUMENTS Federal Specifications Commercial...national or international standardization document developed by a private sector association, organization, or technical society that plans ...Maintain lessons learned • Examples: Guidance for application of a technology; Lists of options Learn. Perform. Succeed. DEFENSE HANDBOOK

  7. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin, E-mail: flamego@ien.gov.b, E-mail: asoares@cnen.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (LIMA/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m{sup 3} of soda almost instantaneously was lost. This inventory contained 1.87x10{sup 10} Bq/m{sup 3} of tritium, 2.22x10{sup 7} Bq/m{sup 3} of cobalt,3.48x10{sup 8} Bq/m{sup 3} of cesium and 3.44x10{sup 10} Bq/m{sup 3} of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10{sup 6} Bq/m{sup 3}, 1,11x10{sup 4} Bq/m{sup 3} and 1,85x10{sup 3} Bq/m{sup 3}) after 22 hours, respectively for {sup 3}H, {sup 60}Co, {sup 131}I and {sup 137}Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  8. Analysis, by Relap5 code, of boron dilution phenomena in a Small Break Loca Transient, performed in PKL III E 2.2 test

    International Nuclear Information System (INIS)

    Rizzo, G.; Vella, G.

    2007-01-01

    The present work is finalized to investigate the E2.2 thermal-hydraulics transient of the PKL III facility, which is a scaled reproduction of a typical German PWR, operated by FRAMATOME-ANP in Erlangen, Germany, within the framework of an international cooperation (OECD/SETH project). The main purpose of the project is to study boron dilution events in Pressurized Water Reactors and to contribute to the assessment of thermal-hydraulic system codes like Relap5. The experimental test PKL III E2.2 investigates the behavior of a typical PWR after a Small Break Loss Of Coolant Accident (SB-LOCA) in a cold leg and an immediate injection of borated water in two cold legs. The main purpose of this work is to simulate the PKL III test facility and particularly its experimental transient by Relap5 system code. The adopted nodalization, already available at Department of Nuclear Engineering (DIN), has been reviewed and applied with an accurate analysis of the experimental test parameters. The main result relies in a good agreement of calculated data with experimental measures for a number of main important variables. (author)

  9. Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

    1985-01-01

    In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

  10. Synergistic behaviour of nuclear radiation, temperature-humidity extremes and LOCA situation on safety and safety-related equipment in Indian nuclear power plants

    International Nuclear Information System (INIS)

    Kulkarni, R.D.; Bora, J.S.; Prakash, Ravi; Agarwal, Vivek; Sundersingh, V.P.

    2002-01-01

    Full text: The general philosophy for the instrumentation in nuclear power plants is based on the use of equipment/instruments which are capable of continuous satisfactory operation over a long period of time with minimum attention. Long term reliability under varying service conditions is of prime importance. The reliability of nuclear power plant depends on the reliability of safety and safety-related electronic instruments/ equipment used for performing the crucial tasks. The electrical and electronic systems/ circuits/ components of the equipment used in reactor safety systems (e.g. reactor protection system, emergency core cooling system, etc.) and reactor safety-related systems (e.g. reactor containment isolation and cooling system, reactor shutdown system, etc.) are responsible for safe and reliable operation of a nuclear power plant. The performance of reactor safety and safety-related equipment/instruments viz. pressure and differential pressure transmitter, amplifier for ion chamber, etc. has been evaluated under synergistic atmosphere including LOCA to find out the critical link in the circuits and subsequent modifications are suggested. The mathematical representation of the generated database has been done to estimate the life span of the instruments and accordingly the guidelines has been prepared for the operational staff to avoid the forced outage of the plant. All the details are included and mathematical models are presented to predict the future performances

  11. Development and Assessment of CFD Models Including a Supplemental Program Code for Analyzing Buoyancy-Driven Flows Through BWR Fuel Assemblies in SFP Complete LOCA Scenarios

    Science.gov (United States)

    Artnak, Edward Joseph, III

    This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-ofcoolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based controlvolume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy.

  12. Comparative severe accident analysis of WWER 1000/B 320 LOCA DN100 computed by computer codes ASTEC V1.1 and SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Kalchev, B.; Dimov, D.; Tusheva, P.; Mladenov, I.

    2005-01-01

    This paper presents the modelling approach for LOCA 100 mm sequence for WWER 1000-B 320 type of reactor with the integral ASTEC computer code and SCDAP/RELAP5 computer code. As a basic input deck the reference input file for Balakovo NPP from the released ASTEC CD has been applied. As a first part of the calculations for the SBLOCA sequence the ASTEC v1.1 modules CESAR, DIVA and CPA have been activated in a coupled mode. For SCDAP/RELAP5 calculation input deck for WWER 1000-B 320 has been applied which meant to be closer to the initial boundary conditions applied for ASTEC WWER 1000 input deck. A SBLOCA 100 mm comparison between ASTEC v1.1 and SCADAP/RELAP5 has been presented. ASTEC predicts vessel failure at 15620 s. ASTEC and SCDAP/RELAP5 give close but not similar results - this could be observed on the trends. The comparison of 100 mm-break shows that SCDAP/RELAP5 predicts clear phenomenological changes in primary pressure evolution and molten pool formation. Similar hydrogen production mass for both codes around 5000 s is detected

  13. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  14. Modelling the transport of radionuclides released in the Ilha Grande bay (Brazil) after a Large Break Loca ion the primary system of a PWR

    International Nuclear Information System (INIS)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Soares, Abner Duarte; Lapa, Celso Marcelo Franklin

    2011-01-01

    It was postulated, in the cooling system of the core, a LOCA, where 431 m 3 of soda almost instantaneously was lost. This inventory contained 1.87x10 10 Bq/m 3 of tritium, 2.22x10 7 Bq/m 3 of cobalt,3.48x10 8 Bq/m 3 of cesium and 3.44x10 10 Bq/m 3 of iodine and was released in liquid form near the Itaorna cove, Angra dos Reis - RJ. Applying the model in the proposed scenario (Angra 1 and 2 in operation and Angra 3 progressively reducing the capture and discharge after the accident), the simulated dilution of the specific activity of radionuclide spots, reached values much lower than report levels for seawater (1,1x10 6 Bq/m 3 , 1,11x10 4 Bq/m 3 and 1,85x10 3 Bq/m 3 ) after 22 hours, respectively for 3 H, 60 Co, 131 I and 137 Cs. From the standpoint of public exposure to radionuclide dispersion, the results of activity concentration obtained by the model suggest that the observed radiological impact is negligible. Based on these findings, we conclude that there would be no radiological impact related to a further release of controlled effluent discharges into Itaorna cove. (author)

  15. Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2009-01-01

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

  16. Simulation with the MELCOR code of two severe accident sequences, Station Blackout and Small Break LOCA, for the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Valle Cepero, Reinaldo

    2004-01-01

    The results of the PSA-I applied to the Atucha I nuclear power plant (CNA I) determine the accidental sequences with the most influence related to the probability of the core reactor damage. Among those sequences are include, the Station Blackout and lost of primary coolant, combine with the failure of the emergency injection systems by pipe breaks of diameters between DN100 - DN25 or equivalent areas, Small LOCA. This paper has the objective to model and analyze the behavior of the primary circuit and the pressure vessel during the evolution of those two accidental sequences. It presented a detailed analysis of the main phenomena that occur from the initial moment of the accident to the failure moment of the pressure vessel and the melt material fall to the reactor cavity. Two sequences were taken into account, considering the main phenomena (core uncover, heating, fuel element oxidation, hydrogen generation, degradation and relocation of the melt material, failure of the support structures, etc.) and the time of occurrence, of those events will be different, if it is considered that both sequences will be developed in different scenarios. One case is an accident with the primary circuit to a high pressure (Station Blackout scenario) and the other with a early primary circuit depressurization due to the lost of primary coolant. For this work the MELCOR 1.8.5 code was used and it allows within a unified framework to modeling an extensive spectrum of phenomenology associated with the severe accidents. (author)

  17. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW) [de

  18. Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

    International Nuclear Information System (INIS)

    Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

    1997-01-01

    The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

  19. BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Yonomoto, Taisuke; Anoda, Yoshinari; Kumamaru, Hiroshige; Nakamura, Hideo; Murata, Hideo; Shiba, Masayoshi; Iriko, Masanori.

    1985-03-01

    This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

  20. Communications standards

    CERN Document Server

    Stokes, A V

    1986-01-01

    Communications Standards deals with the standardization of computer communication networks. This book examines the types of local area networks (LANs) that have been developed and looks at some of the relevant protocols in more detail. The work of Project 802 is briefly discussed, along with a protocol which has developed from one of the LAN standards and is now a de facto standard in one particular area, namely the Manufacturing Automation Protocol (MAP). Factors that affect the usage of networks, such as network management and security, are also considered. This book is divided into three se

  1. Training Standardization

    International Nuclear Information System (INIS)

    Agnihotri, Newal

    2003-01-01

    The article describes the benefits of and required process and recommendations for implementing the standardization of training in the nuclear power industry in the United States and abroad. Current Information and Communication Technologies (ICT) enable training standardization in the nuclear power industry. The delivery of training through the Internet, Intranet and video over IP will facilitate this standardization and bring multiple benefits to the nuclear power industry worldwide. As the amount of available qualified and experienced professionals decreases because of retirements and fewer nuclear engineering institutions, standardized training will help increase the number of available professionals in the industry. Technology will make it possible to use the experience of retired professionals who may be interested in working part-time from a remote location. Well-planned standardized training will prevent a fragmented approach among utilities, and it will save the industry considerable resources in the long run. It will also ensure cost-effective and safe nuclear power plant operation

  2. Effluent standards

    Energy Technology Data Exchange (ETDEWEB)

    Geisler, G C [Pennsylvania State University (United States)

    1974-07-01

    At the conference there was a considerable interest in research reactor standards and effluent standards in particular. On the program, this is demonstrated by the panel discussion on effluents, the paper on argon 41 measured by Sims, and the summary paper by Ringle, et al. on the activities of ANS research reactor standards committee (ANS-15). As a result, a meeting was organized to discuss the proposed ANS standard on research reactor effluents (15.9). This was held on Tuesday evening, was attended by members of the ANS-15 committee who were present at the conference, participants in the panel discussion on the subject, and others interested. Out of this meeting came a number of excellent suggestions for changes which will increase the utility of the standard, and a strong recommendation that the effluent standard (15.9) be combined with the effluent monitoring standard. It is expected that these suggestions and recommendations will be incorporated and a revised draft issued for comment early this summer. (author)

  3. Nuclear standards

    International Nuclear Information System (INIS)

    Fichtner, N.; Becker, K.; Bashir, M.

    1981-01-01

    This compilation of all nuclear standards available to the authors by mid 1980 represents the third, carefully revised edition of a catalogue which was first published in 1975 as EUR 5362. In this third edition several changes have been made. The title has been condensed. The information has again been carefully up-dated, covering all changes regarding status, withdrawal of old standards, new projects, amendments, revisions, splitting of standards into several parts, combination of several standards into one, etc., as available to the authors by mid 1980. The speed with which information travels varies and requires in many cases rather tedious and cumbersome inquiries. Also, the classification scheme has been revised with the goal of better adjustment to changing situations and priorities. Whenever it turned out to be difficult to attribute a standard to a single subject category, multiple listings in all relevant categories have been made. As in previous editions, within the subcategories the standards are arranged by organization (in Categorie 2.1 by country) alphabetically and in ascending numerical order. It covers all relevant areas of power reactors, the fuel cycle, radiation protection, etc., from the basic laws and governmental regulations, regulatory guides, etc., all the way to voluntary industrial standards and codes of pratice. (orig./HP)

  4. Two-phase flow behaviour during a medium size cold leg LOCA test on PMK-2 (IAEA SPE-4)

    International Nuclear Information System (INIS)

    Szabados, L.; Prasser, H.M.; Schaefer, F.

    1995-08-01

    The experiment to the IAEA standard problem exercise No. 4 was carried out in April 1993 on the integral test facility PMK-2 in Budapest. It was a 3.2 mm break on the downcomer hed. The high pressure injection cooling was assumed to be not available. As an accident management measure bleed and feed on the secondary side of the steam generator was applied. Research Center Rossendorf contributed to the experiment of SPE-4 by supplying needle shaped conductivity probes for the measurement of local void fractions in the primary circuit of the PMK-II test facility. In the course of the standard problem exercise No. 4 RCR contributed with posttest calculations using the thermalhydraulic code ATHLET. The report comprises a description of the initial and boundary conditions of the test (chapter 3) and a phenomenological description of the thermalhydraulic events during the transient (chapter 4). The void fraction probe signals yields valuable information for deeper understanding of the thermalhydraulic occurrences and for code validation. This was the case particularly for correction of the level measurement. In chapter 5 a description of the thermalhydraulic occurrences from the viewpoint of code verification is given and the results of RELAP5 (KFKI-AEKI) and of ATHLET-calculations (RCR) are compared. Referring to this description, the sensitivity of the result to main influences is investigated using the ATHLET-code. (orig.) [de

  5. MATE standardization

    Science.gov (United States)

    Farmer, R. E.

    1982-11-01

    The MATE (Modular Automatic Test Equipment) program was developed to combat the proliferation of unique, expensive ATE within the Air Force. MATE incorporates a standard management approach and a standard architecture designed to implement a cradle-to-grave approach to the acquisition of ATE and to significantly reduce the life cycle cost of weapons systems support. These standards are detailed in the MATE Guides. The MATE Guides assist both the Air Force and Industry in implementing the MATE concept, and provide the necessary tools and guidance required for successful acquisition of ATE. The guides also provide the necessary specifications for industry to build MATE-qualifiable equipment. The MATE architecture provides standards for all key interfaces of an ATE system. The MATE approach to the acquisition and management of ATE has been jointly endorsed by the commanders of Air Force Systems Command and Air Force Logistics Command as the way of doing business in the future.

  6. Lugar de encuentros de tópicos románicos : Doña Juana la Loca de Pradilla

    Directory of Open Access Journals (Sweden)

    José Enrique García Melero

    1999-01-01

    Full Text Available Estudio del cuadro titulado «Doña Juana la Loca» de Pradilla, que aquí es considerado como la culminación de la pintura de historia en España y el inicio de su decadencia. Se atribuye su éxito en su época a la concurrencia en él de una serie de características románticas tanto formales como de contenido: la representación de la locura de una Reina enamorada y celosa debido a la muerte de su esposo el Rey Felipe I, tema sacado de la Historia al poco tiempo de acabar la Edad Media y en el inicio del Imperio español. Pero, además, se destacan otros aspectos importantes del cuadro: su escenografía tan teatral y romántica, la representación del amanecer de un día invernal de cielo nublado, el fuego y el humo de una hoguera, el árbol seco, el viento, la figura enlutada de la Reina... Se considera al cuadro como un manierismo del género, como una importante consecuencia del excesivo intelectualismo al que había llegado la pintura de historia hacia 1875. Se establece una relación con el poema de García Lorca titulado «Elegía a Doña Juana la Loca».Learning of the Pradilla's painting titled «Mrs Juana the Lunatic», which is considered here as the culmination of the historical painting in Spain and the commence of its decline. The success of Pradilla's painting in his epoch is attributed to the conjunction in it of a set of characterístics related not only to its forms, but also to its contents: the representation of a Queen who is in love and jealous, and became insane due to the death of her husband, Felipe I. This theme is inspired in the History, little time after the end of the Middle Ages and at the beginning of the Spanish Empire. Besides, there are another aspects of the painting emphasised, such as its theatrical and romantic scenography: the representation of the dawn in a wintry and cloudy day, the fire and smoke caused by a bonfire, the dead tree, the wind, the Queen's in mourning shape... This painting is

  7. A feasibility study on the extended cycle from the point of view of Non-LOCA safety analysis

    International Nuclear Information System (INIS)

    Lee, Chul Sin; Kim, Hee Chul; Kim, Jeong Jin; Baek, Seung Su; Lee, Byung Il; Jeon, Tae Hyun; Lee, Jeong Chan

    1996-06-01

    Extended cycle operation has many advantages as compared with standard cycle operation (12 months) from the viewpoint of operators' exposure to radiation dose, man-power for maintenance and the production of nuclear waste. And it is more economic than the standard cycle operation. If the extended cycle operation is adopted in the CE type nuclear plant, the effective management and enhancement of operation capability could be expected. A feasibility study on the extended cycle operation should be performed. The purpose of this technical report is to perform safety analysis using the design data for Yonggwang Nuclear Plant (YGN) 3 and 4 extended cycle operation and to evaluate qualitatively and quantitatively the safety related design basis events to verify the satisfaction of the safety criteria. Boron dilution and steam line break accidents were found to be most influenced by the change of physics data due to the adaptation of the extended cycle operation. For boron dilution accident, source range monitor ratio of 3.07 for YGN 3 cycle 2 was decreased to 2.92. The 3D reactivity feedback effect due to the local heatup in the vicinity of stuck CEA was credited in the analysis of steam line break. No return-to-power occurred for the steam line break with offsite power available and return-to-power occurred for the steam line break with loss of offsite power. For the steam line break with loss of offsite power, the safety margin was preserved with respect to fuel performance (DNB and LHGR) despite the return-to-power. 6 tabs., 10 figs., 5 refs. (Author) .new

  8. Chemical processes of galvanized steel corrosion in the post-LOCA phase of a PWR and the prevention of sump screen clogging

    International Nuclear Information System (INIS)

    Hoffmann, W.; Kryk, H.

    2012-09-01

    The Emergency Core Coolant System has to remove the decay heat in case of a Loss of Coolant Accident (LOCA). Therefore, the emergency core cooling pumps recirculate the fluid from the sump back into the primary circuit. Sump strainers are mounted at the pump inlets to retain particles and fibrous insulation material. A fiber bed formed on strainers may act as an additional debris filter. However, a critical increase of pressure drop generated by debris or corrosion products could cause a failure of emergency cooling. Problems of insulation materials NUKON R (fiberglass) or CalSil and Aluminium may appear if containment spray systems using alkaline additives are installed. In such cases, dissolution / precipitation reactions resulting from insulation materials were observed, which increase the risk of sump screen blockage. In German NPPs, there are no containments spray systems, and insulation consists of more resistant materials like mineral wool (rock wool) and stainless steel. However, large scale experiments from AREVA have shown that sump screen clogging may be initiated by boric acid containing For generic investigations of galvanized steel corrosion behaviour under post-LOCA conditions, the down-scaled test facility KorrVA was designed consisting of a loop with trickle section (location of LOCA), bath section (sump), horizontal strainer and circulation pump. The low coolant volume (60 L) permits an easy and efficient purification between the experiments including complete removal of corrosion products. About 90 experiments were carried out with galvanized steel gratings and galvanized steel coupons in boric acid media in order to determine corrosion mechanisms depending on different experimental conditions like temperature, water chemistry and hydrodynamic conditions (flow impact, simulated by different nozzles). Practically, the fiber bed was prepared during a preliminary stage with the aim to separate effects of fiber bed formation on sump strainer clogging

  9. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2010-02-01

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

  10. Frequency standards

    CERN Document Server

    Riehle, Fritz

    2006-01-01

    Of all measurement units, frequency is the one that may be determined with the highest degree of accuracy. It equally allows precise measurements of other physical and technical quantities, whenever they can be measured in terms of frequency.This volume covers the central methods and techniques relevant for frequency standards developed in physics, electronics, quantum electronics, and statistics. After a review of the basic principles, the book looks at the realisation of commonly used components. It then continues with the description and characterisation of important frequency standards

  11. The case for improved HEPA-filter mechanical performance standards revisited

    Energy Technology Data Exchange (ETDEWEB)

    Ricketts, C.I.; Smith, P.R. [New Mexico State Univ., Las Cruces, NM (United States)

    1997-08-01

    Under benign operating conditions, High Efficiency Particulate Air (HEPA) filter units serve as reliable and relatively economical components in the air cleaning systems of nuclear facilities worldwide. Despite more than four decades of filter-unit evaluation and improvements, however, the material strength characteristics of the glass fiber filter medium continue to ultimately limit filter functional reliability. In worst-case scenarios involving fire suppression, loss-of-coolant accidents (LOCA`s), or exposure to shock waves or tornado induced flows, rupture of the filter medium of units meeting current qualification standards cannot be entirely ruled out. Even under so-called normal conditions of operation, instances of filter failure reported in the literature leave open questions of filter-unit reliability. Though developments of filter units with improved burst strengths have been pursued outside the United States, support for efforts in this country has been comparatively minimal. This despite user requests for filters with greater moisture resistance, for example. Or the fact that conventional filter designs result in not only the least robust component to be found in a nuclear air cleaning system, but also the one most sensitive to the adverse effects of conditions deviating from those of normal operation. Filter qualification-test specifications of current codes, standards, and regulatory guidelines in the United States are based primarily upon research performed in a 30-year period beginning in the 1950`s. They do not seem to reflect the benefits of the more significant developments and understanding of filter failure modes and mechanisms achieved since that time. One overseas design, based on such knowledge, has proven reliability under adverse operating conditions involving combined and serial challenges. Its widespread use, however, has faltered on a lack of consensus in upgrading filter performance standards. 34 refs., 2 figs., 3 tabs.

  12. Relevant Standards

    Indian Academy of Sciences (India)

    .86: Ethernet over LAPS. Standard in China and India. G.7041: Generic Framing Procedure (GFP). Supports Ethernet as well as other data formats (e.g., Fibre Channel); Protocol of ... IEEE 802.3x for flow control of incoming Ethernet data ...

  13. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2014-01-01

    International e-Customs is going through a standardization process. Driven by the need to increase control in the trade process to address security challenges stemming from threats of terrorists, diseases, and counterfeit products, and to lower the administrative burdens on traders to stay...

  14. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2016-01-01

    International e-Customs is going through a standardization process. Driven by the need to increase control in the trade process to address security challenges stemming from threats of terrorists, diseases, and counterfeit products, and to lower the administrative burdens on traders to stay...

  15. Standard Fortran

    International Nuclear Information System (INIS)

    Marshall, N.H.

    1981-01-01

    Because of its vast software investment in Fortran programs, the nuclear community has an inherent interest in the evolution of Fortran. This paper reviews the impact of the new Fortran 77 standard and discusses the projected changes which can be expected in the future

  16. An experimental study of the corrosion and precipitation of aluminum in the presence of trisodium phosphate buffer following a loss of coolant accident (LOCA) scenario

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Jun [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry J. [Department of Civil Engineering, University of New Mexico (United States); Leavitt, Janet J. [Department of Civil Engineering, University of New Mexico (United States); Alion Science and Technology (United States); Hammond, Kyle; Mitchell, Lana [Department of Civil Engineering, University of New Mexico (United States); Kee, Ernie [South Texas Project Nuclear Operating Company (STPNOC) (United States); Blandford, Edward D., E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States)

    2015-02-15

    Highlights: • Experimental head loss testing was conducted by aggressively promoting corrosion in loss of coolant accidents. • Blender-processed debris beds have higher head loss but tend to be less reproducible than NEI-processed debris beds. • Precipitation was observed from aluminum concentration and turbidity measurements. • Precipitation results were compared to predictions from Visual MINTEQ. - Abstract: This paper presents the results of an integrated chemical effects experiment of head loss across the sump pump screen with fibrous debris bed over a non-prototypical 10-day post-LOCA incident window. The corrosion head loss experiments (CHLE) is a reduced scaled integral effects testing facility built at the University of New Mexico (UNM) to investigate potential chemical effects on head loss across prepared fibrous debris beds. The results in this paper come from two integral effect tests performed at UNM in order to determine the chemical effects on head loss induced by a zinc source effect and an aluminum precipitation effect (T3: without Zn source case, T4: with Zn source case in containment). The tests were performed with a large surface area of aluminum coupons in the testing facility for an extended period of elevated temperature to accelerate corrosion above that expected under prototypical conditions. These conditions were sufficient to force aluminum precipitation to occur and induce the onset of chemical effects on debris bed head loss. The head loss behavior on two different types of fiber debris beds (blender-processed and NEI-processed debris bed) was evaluated in this study. It was found that the blender-processed bed is much more sensitive in filtering than the NEI-processed bed and consequently had a much higher head loss value across the beds. Aluminum precipitation was observed, with aluminum concentration and turbidity measurements, to form starting on day 7 in Test T3 and on day 6 in Test T4. The onset of aluminum precipitation

  17. Modelo para o dimensionamento e a locação de sistemas de terraceamento em nível Model to design level terracing systems

    Directory of Open Access Journals (Sweden)

    Nori P. Griebeler

    2005-12-01

    Full Text Available A erosão hídrica é um dos principais causadores da degradação das terras agrícolas, provocando expressivos prejuízos à sociedade. Em virtude desses danos, torna-se fundamental o desenvolvimento de técnicas que permitam o aumento da eficiência das práticas para a conservação do solo. Com esse objetivo, desenvolveu-se um modelo para o dimensionamento e a locação de sistemas de terraceamento em nível, utilizando modelos digitais de elevação do terreno gerados em sistemas de informações geográficas (SIG. A fim de possibilitar a aplicação do modelo desenvolvido, elaborou-se um software que permite a realização das interações necessárias para o dimensionamento do sistema. Os resultados obtidos mostram que o modelo oferece a vantagem de permitir o dimensionamento de cada terraço, considerando as condições existentes na área específica que irá contribuir com o escoamento superficial para o mesmo. A outra vantagem do modelo desenvolvido está na possibilidade de realização de simulações, permitindo a observação e o acompanhamento, de maneira fácil e objetiva, das mudanças provocadas no sistema de terraceamento, em função da alteração nas características de uso e manejo do solo.Water erosion is one of the main problems of agricultural lands degradation, causing largess damages to the society. Because of these damages, it becomes necessary the development of techniques that allow the increase of the efficiency of soil conservation practices. With this objective, a model was developed for design and location of level terracing systems using digital models of elevation of the land generated in geographic information systems (GIS. In order to facilitate the application of the developed model, it was elaborated a software. The results show that the model offers as the principal advantage the design of each terrace considering the existent conditions in the specific area that will contribute with the runoff to

  18. An experimental study of the corrosion and precipitation of aluminum in the presence of trisodium phosphate buffer following a loss of coolant accident (LOCA) scenario

    International Nuclear Information System (INIS)

    Kim, Seung Jun; Howe, Kerry J.; Leavitt, Janet J.; Hammond, Kyle; Mitchell, Lana; Kee, Ernie; Blandford, Edward D.

    2015-01-01

    Highlights: • Experimental head loss testing was conducted by aggressively promoting corrosion in loss of coolant accidents. • Blender-processed debris beds have higher head loss but tend to be less reproducible than NEI-processed debris beds. • Precipitation was observed from aluminum concentration and turbidity measurements. • Precipitation results were compared to predictions from Visual MINTEQ. - Abstract: This paper presents the results of an integrated chemical effects experiment of head loss across the sump pump screen with fibrous debris bed over a non-prototypical 10-day post-LOCA incident window. The corrosion head loss experiments (CHLE) is a reduced scaled integral effects testing facility built at the University of New Mexico (UNM) to investigate potential chemical effects on head loss across prepared fibrous debris beds. The results in this paper come from two integral effect tests performed at UNM in order to determine the chemical effects on head loss induced by a zinc source effect and an aluminum precipitation effect (T3: without Zn source case, T4: with Zn source case in containment). The tests were performed with a large surface area of aluminum coupons in the testing facility for an extended period of elevated temperature to accelerate corrosion above that expected under prototypical conditions. These conditions were sufficient to force aluminum precipitation to occur and induce the onset of chemical effects on debris bed head loss. The head loss behavior on two different types of fiber debris beds (blender-processed and NEI-processed debris bed) was evaluated in this study. It was found that the blender-processed bed is much more sensitive in filtering than the NEI-processed bed and consequently had a much higher head loss value across the beds. Aluminum precipitation was observed, with aluminum concentration and turbidity measurements, to form starting on day 7 in Test T3 and on day 6 in Test T4. The onset of aluminum precipitation

  19. The System 80+ Standard Plant design control document. Volume 23

    International Nuclear Information System (INIS)

    1997-01-01

    This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains part 16 References and Appendix 19 A Design Alternatives for section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Also covered is section 20 Unresolved Safety Issues of the ADM Design and Analysis. Finally sections 1--6 of the ADM Emergency Operations Guidelines are contained in this volume. Information covered in these sections include: standard post-trip actions; diagnostic actions; reactor trip recovery guideline; LOCA recovery; SG tube rupture recovery

  20. Condensation During Nuclear Reactor Loca

    International Nuclear Information System (INIS)

    Rihan, Y.; Teamah, M.; Sorour, M.; Soliman, S.

    2008-01-01

    Two-phase channel flow with condensation is a common phenomenon occurs in a number of nuclear reactor accident scenarios. It also plays an important role during the operation of the safety coolant injection systems in advanced nuclear reactors. Semiempirical correlations and simple models based on the analogy between heat and mass transfer processes have been previously applied. Rigorous models, compatible with the state-of-the-art numerical algorithms used in thermal-hydraulic computer codes, are scare, and are of great interest. The objective of this research is to develop a method for modeling condensation, with noncondensable gases, compatible with the state-of-the-art numerical methods for the solution of multi-phase field equations. A methodology for modeling condensation, based on the stagnant film theory, and compatible with the reviewed numerical algorithms, is developed. The model treats the coupling between the heat and mass transfer processes, and allows for an implicit treatment of the mass and momentum exchange terms as the gas-liquid interphase, without iterations. The developed model was used in the application of loss of coolant in pressurized water reactor accidents

  1. LOCA verification and data bank

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.; Cox, N.D.; Atwood, C.L.; Madden, S.C.; Condie, K.G.

    1979-01-01

    The purpose of this task was to derive local conditions heat transfer parameters and their uncertainties using computer codes and experimentally derived boundary conditions. To accomplish this objective, Semiscale S-02-9 blowdown experiment was used along with the INVERT (an inverse heat conduction code) and RELAP4 (a thermal hydraulic code) codes as the analytical tools. The uncertainties calculated for the local conditions were limited to those introduced by inaccuracies in the experimentally measured boundary conditions. The propagation of the measurement uncertainties through the codes was investigated by varying the code input using statistical methods and a response surface technique

  2. Advanced LOCA code uncertainty assessment

    International Nuclear Information System (INIS)

    Wickett, A.J.; Neill, A.P.

    1990-11-01

    This report describes a pilot study that identified, quantified and combined uncertainties for the LOBI BL-02 3% small break test. A ''dials'' version of TRAC-PF1/MOD1, called TRAC-F, was used. (author)

  3. Importance of water quality on plant abundance and diversity in high-alpine meadows of the Yerba Loca Natural Sanctuary at the Andes of north-central Chile Importancia de la calidad del agua sobre la abundancia y diversidad vegetal en vegas altoandinas del Santuario Natural Yerba Loca en los Andes de Chile centro-norte

    Directory of Open Access Journals (Sweden)

    ROSANNA GINOCCHIO

    2008-12-01

    Full Text Available Porphyry Cu-Mo deposits have influenced surface water quality in high-Andes of north-central Chile since the Miocene. Water anomalies may reduce species abundance and diversity in alpine meadows as acidic and metal-rich waters are highly toxic to plants The study assessed the importance of surface water quality on plant abundance and diversity in high-alpine meadows at the Yerba Loca Natural Santuary (YLNS, central Chile (33°15' S, 70°18' W. Hydrochemical and plant prospecting were carried out on Piedra Carvajal, Chorrillos del Plomo and La Lata meadows the growing seasons of 2006 and 2007. Direct gradient analysis was performed through canonical correspondence analysis (CCA to look for relationships among water chemistry and plant factors. High variability in water chemistry was found inside and among meadows, particularly for pH, sulphate, electric conductivity, hardness, and total dissolved Cu, Zn, Cd, Pb and Fe. Data on species abundance and water chemical factors suggests that pH and total dissolved Cu are very important factor determining changes in plant abundance and diversity in study meadows. For instance, Festuca purpurascens, Colobanthus quitensis, and Arenaria rivularis are abundant in habitals with Cu-rich waters while Festuca magellanica, Patosia clandestina, Plantago barbata, Werneria pygmea, and Erigeron andícola are abundant in habitals with dilute waters.Los megadepósitos de pórfidos de Cu-Mo han influido sobre la calidad de las aguas superficiales en las zonas altoandinas del centro-norte de Chile desde el Mioceno. Estas alteraciones en la calidad de las aguas podrían afectar negativamente a la vegetación presente en las vegas altoandinas, ya que las aguas acidas y ricas en metales son altamente tóxicas para las plantas. En este estudio se evaluó el efecto de la calidad de las aguas en la abundancia y diversidad florística de las vegas altoandinas del Santuario de la Naturaleza Yerba Loca (SNYL, en Chile central (33

  4. The CSNI International Standard Problem Programme: Overall Presentation on Objectives; Rationale and Lessons Learnt: a Joint Venture of the Thermalhydraulic International Community

    International Nuclear Information System (INIS)

    Reocreux, M.

    2008-01-01

    The CSNI International Standard Problems have been one of the key activities of the CSNI thermal hydraulics groups during the last 25 years. After recalling the way the international standard problems were initiated in the late 1970 years -they were called at that time CSNI LOCA Standard Problem- the process which has led to make from the ISPs a full CSNI activity, is described. Rules have been defined which formalized the way experimental results were provided and the way the comparison exercise were performed. The long series of ISPs from 1975 up to nowadays is described, explaining the different trends in the ISPs choices. The findings which have been obtained are reviewed on both technical and programmatic aspects.

  5. Nova abordagem metodológica para locação, modelagem 3D e monitoramento de barragens subterrâneas no semiárido brasileiro

    OpenAIRE

    Lima, Alexandre de Oliveira

    2013-01-01

    Apesar do número significativo de barragens construídas nos últimos anos no Brasil e mesmo em outros países, são poucos os estudos técnicos que dão suporte a este crescimento. Com o objetivo de propor uma nova abordagem metodológica para o processo de locação e monitoramento de barragens subterrâneas, em condições de clima semiárido, foram realizados estudos em duas sub-bacias hidrográficas (do rio das Cobras e Pau Lavrado) ambas situadas no semiárido do Estado do Rio Grande. Nesta nova abord...

  6. The Dynamics of Standardization

    DEFF Research Database (Denmark)

    Brunsson, Nils; Rasche, Andreas; Seidl, David

    2012-01-01

    This paper suggests that when the phenomenon of standards and standardization is examined from the perspective of organization studies, three aspects stand out: the standardization of organizations, standardization by organizations and standardization as (a form of) organization. Following a comp...

  7. Conclusions Drawn from the Investigation of LOCA-Induced Insulation Debris Generation and its Impact on Emergency Core Cooling (ECC) at German NPPs - Approach Taken by / Perspective of The German TSO (TuV)

    International Nuclear Information System (INIS)

    Huber, J.

    2004-01-01

    Initiated by the Barsebaeck incident in 1992 and the following activities related to the LOCA-induced insulation debris generation and its impact on emergency core cooling, investigations on German PWRs and BWRs were performed in these areas. The investigations on the German BWRs were carried out in detail immediately after the Barsebaeck incident in the years 1992 through 1994. Detailed investigations on the German PWRs started after the issue of the OECD report in 1996. Therefore the investigations on the impact of LOCA-induced insulation debris generation on strainer plugging carried out in Germany in the last years were focused mainly on the German PWRs. In the framework of these investigations of the German PWRs, the relevant parameters and phenomena were investigated in detail by the plant owners in the years from 1997 through 1999. The results were summarised in reports for each plant. The main results of the investigations conducted by the plant owners were that the plant owners considered backfitting in German PWRs is not necessary to guarantee emergency core cooling following a LOCA with insulation debris generation. As the technical support organisation for the German Bavarian and Hessian state authority, the TUV Suddeutschland was called upon to examine these investigations and the conclusions drawn by the plant owners. We compared each of the parameters and phenomena against the state of knowledge. The results of our examination in 1999 showed that the investigations of the plant owners were generally correct, but we stated also, that due to existing uncertainties, further investigations are necessary to validate the results. To meet these demands, the plant owners installed a working group for planning and performing newer, more realistic large-scale experiments (scaling factor 1:4) to investigate the transport mechanism of the insulation material within the containment sump, the head loss at the strainers and to estimate the amount of insulation

  8. Standards for Standardized Logistic Regression Coefficients

    Science.gov (United States)

    Menard, Scott

    2011-01-01

    Standardized coefficients in logistic regression analysis have the same utility as standardized coefficients in linear regression analysis. Although there has been no consensus on the best way to construct standardized logistic regression coefficients, there is now sufficient evidence to suggest a single best approach to the construction of a…

  9. Instant standard concept for data standards development

    NARCIS (Netherlands)

    Folmer, Erwin Johan Albert; Kulcsor, Istvan Zsolt; Roes, Jasper

    2013-01-01

    This paper presents the current results of an ongoing research about a new data standards development concept. The concept is called Instant Standard referring to the pressure that is generated by shrinking the length of the standardization process. Based on this concept it is estimated that the

  10. Malaysian NDT standards

    International Nuclear Information System (INIS)

    Khazali Mohd Zin

    2001-01-01

    In order to become a developed country, Malaysia needs to develop her own national standards. It has been projected that by the year 2020, Malaysia requires about 8,000 standards (Department of Standard Malaysia). Currently more than 2,000 Malaysian Standards have been gazette by the government which considerably too low before tire year 2020. NDT standards have been identified by the standard working group as one of the areas to promote our national standards. In this paper the author describes the steps taken to establish the Malaysian very own NDT standards. The project starts with the establishment of radiographic standards. (Author)

  11. Selva, plumas y desconche : Un análisis de las performances masculinas de la feminidad entre las locas del Tigre durante la década del ochenta.

    Directory of Open Access Journals (Sweden)

    Santiago Joaquin Insausti

    2011-08-01

    Full Text Available En este trabajo me propongo describir los modos en que las personas que ejercían géneros o sexualidades diversas construían sus subjetividades y sus cuerpos en Argentina en las décadas del setenta y el ochenta. Para esto, partiré del análisis de las fotos que estas personas se tomaban en el contexto de fiestas clandestinas organizadas en las islas del Tigre. Partiendo de la tesis constructivista de la ausencia de distancia entre género y expresión de género, indagare en las fotos la relación de las imágenes con el contexto selvático que las enmarca, la construcción contrapuesta de chongos y locas, dos de las principales identidades sexuales de la época, y en los usos del cuerpo utilizados por estos para producir feminidad y masculinidad. Finalmente, arriesgare dos hipótesis tendientes a explicar las causas de la masiva migración de estos sujetos desde la ciudad hacia el delta del Paraná.

  12. Selva, plumas y desconche : Un análisis de las performances masculinas de la feminidad entre las locas del Tigre durante la década del ochenta.

    Directory of Open Access Journals (Sweden)

    Santiago Joaquin Insausti

    2010-08-01

    Full Text Available En este trabajo me propongo describir los modos en que las personas que ejercían géneros o sexualidades diversas construían sus subjetividades y sus cuerpos en Argentina en las décadas del setenta y el ochenta. Para esto, partiré del análisis de las fotos que estas personas se tomaban en el contexto de fiestas clandestinas organizadas en las islas del Tigre. Partiendo de la tesis constructivista de la ausencia de distancia entre género y expresión de género, indagare en las fotos la relación de las imágenes con el contexto selvático que las enmarca, la construcción contrapuesta de chongos y locas, dos de las principales identidades sexuales de la época, y en los usos del cuerpo utilizados por estos para producir feminidad y masculinidad. Finalmente, arriesgare dos hipótesis tendientes a explicar las causas de la masiva migración de estos sujetos desde la ciudad hacia el delta del Paraná.