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.

Tools evaluation and development for loss of coolant accidents analysis in research reactors

International Nuclear Information System (INIS)

Maprelian, Eduardo; Cabral, Eduardo L.L.; Silva, Antonio T. e

1999-01-01

The loss of coolant accidents (LOCA) in pool type research reactors are normally considered as limiting in the licensing process. This paper verifies the viability of the computer code 3D-AIRLOCA to analyze LOCA in a pool type research reactor, and also develops two computer codes LOSS and TEMPLOCA. The computer code LOSS determines the time tom drawn the pool down to the level of the bottom of the core, and the computer code TEMPLOCA calculates the peak fuel element temperature during the transient. These two coders substitutes the 3D-AIRLOCA in the LOCA analysis for pool type research reactors. (author)

Alternative protections for loss of coolant accidents

International Nuclear Information System (INIS)

Estevez, E.A.

1997-01-01

One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

Definition of loss-of-coolant accident radiation source. [PWR; BWR

Energy Technology Data Exchange (ETDEWEB)

1978-02-01

Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

Consequences in the pumps operation during a large loss of coolant accident

International Nuclear Information System (INIS)

Santos, G.A. dos; Sabundjian, G.

1991-08-01

The event of living on or turning off the operation of the Reactor Cooling Pumps - RCPs, in the case of a Loss of Coolant Accident - LOCA, has been a reason of a lot of studies after the Three Mile Island 2 accident. Thus, it was investigated a large break LOCA in the cold leg of Angra 1, with the RELAP4/MOD5 Code during the blowdown. The attained results indicated that the best performance of the core was in the case where the RCPs had been turned off in the beginning of the transient, when compared with different operation conditions of the RCPs. (author)

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)

Research on loss of coolant accident of pressurized-water reactor based on PSO algorithm

International Nuclear Information System (INIS)

Ma Jie; Guo Lifeng; Peng Qiao

2012-01-01

In order to improve the diagnosis performance of Loss of Coolant Accident (LOCA), based on Back Propagation (BP) algorithm study, a fault diagnosis network is established based on Particle Swarm Optimization (PSO) algorithm in this paper. The PSO algorithm is used to train the weights and the thresholds of neural network, which can conquer part convergence problem of BP algorithm. The test results show that the diagnosis network has higher accuracy of LOCA. (authors)

Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-01-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1994-07-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

International Nuclear Information System (INIS)

Abass, N. M. N.

2012-02-01

A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

International Nuclear Information System (INIS)

Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul

2015-01-01

Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree

Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

Energy Technology Data Exchange (ETDEWEB)

Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

2015-05-15

Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

Small break LOCA [loss of coolant accident] mitigation for Bellefonte

International Nuclear Information System (INIS)

Bayless, P.D.; Dobbe, C.A.

1986-01-01

Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

Analysis of forces on core structures during a loss-of-coolant accident. Final report

International Nuclear Information System (INIS)

Griggs, D.P.; Vilim, R.B.; Wang, C.H.; Meyer, J.E.

1980-08-01

There are several design requirements related to the emergency core cooling which would follow a hypothetical loss-of-coolant accident (LOCA). One of these requirements is that the core must retain a coolable geometry throughout the accident. A possible cause of core damage leading to an uncoolable geometry is the action of forces on the core and associated support structures during the very early (blowdown) stage of the LOCA. An equally unsatisfactory design result would occur if calculated deformations and failures were so extensive that the geometry used for calculating the next stages of the LOCA (refill and reflood) could not be known reasonably well. Subsidiary questions involve damage preventing the operation of control assemblies and loss of integrity of other needed safety systems. A reliable method of calculating these forces is therefore an important part of LOCA analysis. These concerns provided the motivation for the study. The general objective of the study was to review the state-of-the-art in LOCA force determination. Specific objectives were: (1) determine state-of-the-art by reviewing current (and projected near future) techniques for LOCA force determination, and (2) consider each of the major assumptions involved in force determination and make a qualitative assessment of their validity

Use of flow models to analyse loss of coolant accidents

International Nuclear Information System (INIS)

Pinet, Bernard

1978-01-01

This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

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)

Source term and behavioural parameters for a postulated HIFAR loss-of-coolant accident

International Nuclear Information System (INIS)

May, F.G.

1987-01-01

The fraction of the fission product inventory which might be released into the atmosphere of the HIFAR reactor containment building (RCB) during a postulated loss-of-coolant accident (LOCA) has been evaluated as a function of time, for each classification of airborne radioactivity. This appraisal will be used as the source term for a computer program, which uses realistic attenuation of the fission product aerosol in a single compartment model with a defined leakrate to predict possible radioactive releases into the environment in a hypothetical bounding case reactor accident which is rather more severe in all major aspects than any single LOCA. Also given are the parameters governing the attenuation of the aerosol and vapours in the atmosphere of the RCB so that their behaviour may be accurately modelled. The source terms for several other types of accident involving the meltdown of fuel elements have also been considered but in less detail than the LOCA case. In some of the cases, the fission products are released directly to atmosphere, so there is no attenuation of the release by deposition within the RCB

ESBWR long term containment response to loss of coolant accidents

International Nuclear Information System (INIS)

Alamgir, M. D.; Marquino, W.; Diaz-Quiroz, J.; Tucker, L.

2010-01-01

ESBWR is a 4500 MWt generation III+ natural circulation reactor with an array of robust passive safety systems to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US certification process. This paper focuses on the bounding licensing analysis of the long-term (30-day) response of the ESBWR containment to limiting Loss of Coolant Accident (LOCA) performed with the TRACG code. It is shown that using only passive systems available during the first 72 hours after the limiting Main Steam Line Break LOCA, the predicted peak containment pressure in the ESBWR containment remain well below the design limits with good margin. After 72 hours of LOCA initiation, PCCS Vent Fans (non-safety system) become available that remove non-condensable gases from, and further enhance the effectiveness of, PCCS heat exchangers to reduce the containment pressure and temperature to values substantially below the design limits. During the post- 72 hour period, the beneficial effects of the Vent Fan operation, combined with the available operator action to refill of PCCS pools, continue to maintain the containment pressure to about 30% below the design limit at 30 days after a limiting ESBWR LOCA. (authors)

Loss-of-coolant accident for large pipe breaks in light water reactor plants

International Nuclear Information System (INIS)

Keusenhoff, J.

1980-01-01

The importance of loss-of-coolant accidents (LOCA) and their control for nuclear reactor safety is explained. Showing the cooling circuits and emergency core cooling systems (ECCS) of both, PWR and BWR, the possible break spectrum and the general sequence of events is discussed. The governing physical phenomena for the different LOCA phases are pointed out in more detail. Special emphasis is taken on rules, regulations and failure criteria for licensing purposes. Analysis methods and codes for both, evaluation and best-estimate model are compared under deterministic and probabilistic approach, respectively. Some insight in present integral and separate effect tests demonstrates the interdependency of analysis and experiment. Results of LOCA analysis and experiments show the present state of the art. (orig.)

Generic evaluation of small break loss-of-coolant accident behavior in Babcock and Wilcox designed 177-FA operating plants

International Nuclear Information System (INIS)

1980-01-01

Slow system depressurization resulting from small break loss-of-coolant accidents (LOCAs) in the reactor coolant system have not, until recently, received detailed analytical study comparable to that devoted to large breaks. Following the TMI-2 accident, the staff had a series of meetings with Babcock and Wilcox (B and W) and the B and W licensees. The staff requested that B and W and the licensees: (1) systematically evaluate plant response for small break loss-of-coolant accidents; (2) address each of the concerns documented in the Michelson report; (3) validate the computer codes used against the TMI-2 accident; (4) extend the break spectrum analysis to very small breaks, giving special consideration to failure of pressurizer valves to close; (5) analyze degraded conditions where AFW is not available; (6) prepare design changes aimed at reducing the probability of loss-of-coolant accidents produced by the failure of a PORV to close; and (7) develop revised emergency procedures for small breaks. This report describes the review of the generic analyses performed by B and W based on the requests stated above

Pressure behavior in nuclear reactor containment following a loss of coolant accident

Energy Technology Data Exchange (ETDEWEB)

Khattab, M; Ibrahim, N A; Bedrose, C D [Reactors department, nuclear research center, atomic energy authority, Cairo, (Egypt)

1995-10-01

The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs.

Pressure behavior in nuclear reactor containment following a loss of coolant accident

International Nuclear Information System (INIS)

Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

1995-01-01

The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

Pressure behaviour in a nuclear reactor containment following a loss of coolant accident

International Nuclear Information System (INIS)

KHattab, M.S.; Ibrahim, N.A.; Bedrose, S.D.

1994-01-01

The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break, is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The scenarios of small, medium and large LOCA at 2%, 15% and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The results of large LOCA showed good agreement with westinghouse calculations of the same design. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs., 1 tab

Identification of Loss-of-Coolant Accidents in LWRs by Inverse Models

International Nuclear Information System (INIS)

Cholewa, Wojciech; Frid, Wiktor; Bednarski, Marcin

2004-01-01

This paper describes a novel diagnostic method based on inverse models that could be applied to identification of transients and accidents in nuclear power plants. In particular, it is shown that such models could be successfully applied to identification of loss-of-coolant accidents (LOCAs). This is demonstrated for LOCA scenarios for a boiling water reactor. Two classes of inverse models are discussed: local models valid only in a selected neighborhood of an unknown element in the data set, representing a state of a considered object, and global models, in the form of partially unilateral models, valid over the whole learning data set. An interesting and useful property of local inverse models is that they can be considered as example-based models, i.e., models that are spanned on particular sets of pattern data. It is concluded that the optimal diagnostic method should combine the advantages of both models, i.e., the high quality of results obtained from a local inverse model and the information about the confidence interval for the expected output provided by a partially unilateral model

Computer programmes of the Power Research Institute for the analysis of processes in the primary coolant circuit and in the containment of a WWER plant in a loss-of-coolant accident

International Nuclear Information System (INIS)

Misak, J.

1976-01-01

A brief description is given of computer programmes for the analysis of loss-of-coolant accidents (LOCA) in WWER type reactors. The LENKA programme is intended for the thermal and hydraulic analysis of the consequences of such accidents in the primary coolant circuit. The SICHTA programme is intended for the detailed calculation of the time dependence of the axial and radial distribution of heat in fuel rods from steady-state to the flooding of the core. CHEMLOC is intended for the analysis of the heat history of the core and the extent of chemical reactions in LOCA when the emergency core cooling system is not operating. The TRACO I is intended for the analysis of the initial stage of the transient process in a full-pressure containment after LOCA (the computation of the time and spatial dependences of pressures and temperatures). TRACO III is intended for the computation of the long-term time dependence of pressure and temperature in the full-pressure containment after LOCA. (B.S.)

Thermal-hydraulic analysis of loss-of-coolant accident in the JMTR

International Nuclear Information System (INIS)

Sakurai, Fumio; Oyamada, Rokuro

1985-02-01

The reevaluation of the Loss-of-Coolant Accident (LOCA) was required through the process of a safety review for the Japan Materials Testing Reactor (JMTR) core conversion from the high-enriched uranium fuel (Enrichment : 93%) to the medium-enriched uranium fuel (Enrichment : 45%). The following were concluded by thermal-hydraulic analysis of a LOCA caused by a double-ended pipe break in the JMTR primary cooling system. (1) The fuel in the core does not burn-out as long as it is covered with water. (2) A larger siphon break valve (larger than phi60mm) should be installed instead of the present one (phi25mm) on the primary cooling system in order to prevent the core from being uncovered with water in case of a LOCA caused by a double-ended pipe break. The present siphon break valve was installed to keep the core covered with water in case of a LOCA caused by a small pipe rupture. In this analysis, the Siphon Breaker Analysis Code (SBAC) was written in order to analyse the size of the siphon break valve and its accuracy was confirmed to be within 5% through a verification experiment. (author)

Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

International Nuclear Information System (INIS)

Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

1995-01-01

The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

International Nuclear Information System (INIS)

Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

1998-04-01

For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs

Analysis of loss of coolant accident and emergency core cooling system

International Nuclear Information System (INIS)

Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

1977-01-01

In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

1995-01-01

Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

Cladding embrittlement during postulated loss-of-coolant accidents.

Energy Technology Data Exchange (ETDEWEB)

Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

2008-07-31

The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

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

Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

International Nuclear Information System (INIS)

Nguyen Van Thai; Kieu Ngoc Dung

2015-01-01

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

Estimation of break location and size for loss of coolant accidents using neural networks

International Nuclear Information System (INIS)

Na, Man Gyun; Shin, Sun Ho; Jung, Dong Won; Kim, Soong Pyung; Jeong, Ji Hwan; Lee, Byung Chul

2004-01-01

In this work, a probabilistic neural network (PNN) that has been applied well to the classification problems is used in order to identify the break locations of loss of coolant accidents (LOCA) such as hot-leg, cold-leg and steam generator tubes. Also, a fuzzy neural network (FNN) is designed to estimate the break size. The inputs to PNN and FNN are time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. An automatic structure constructor for the fuzzy neural network automatically selects the input variables from the time-integrated values of many measured signals, and optimizes the number of rules and its related parameters. It is verified that the proposed algorithm identifies very well the break locations of LOCAs and also, estimate their break size accurately

Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

International Nuclear Information System (INIS)

Pitts, J.H.; McCauley, E.W.

1978-01-01

The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

Loss of coolant accident mitigation for liquid metal cooled space reactors

International Nuclear Information System (INIS)

Georgevich, Vladimir; Best, Frederick; Erdman, Carl

1989-01-01

A loss of coolant accident (LOCA) in a liquid metal-cooled space reactor system has been considered as a possible accident scenario. Development of new concepts that will prevent core damage by LOCA caused elevated temperatures is the primary motivation of this work. Decay heat generated by the fission products in the reactor core following shutdown is sufficiently high to melt the fuel unless energy can be removed from the pins at a sufficiently rapid rate. There are two major reasons that prevent utilization of traditional emergency cooling methods. One is the absence of gravity and the other is the vacuum condition outside the reactor vessel. A concept that overcomes both problems is the Saturated Wick Evaporation Method (SWEM). This method involves placing wicking structures at specific locations in the core to act as energy sinks. One of its properties is the isothermal behaviour of the liquid in the wick. The absorption of energy by the surface at the isothermal temperature will direct the energy into an evaporation process and not in sensible heat addition. The use of this concept enables establishment of isothermal positions within the core. A computer code that evaluates the temperature distribution of the core has been developed and the results show that this design will prevent fuel meltdown. (author)

Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1991-01-01

The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

CNE (central nuclear en Embalse): probabilistic safety study. Loss-of-coolant accidents. Analysis through events sequence

International Nuclear Information System (INIS)

Layral, S.I.

1987-01-01

The aim of this study was to perform for the Embalse nuclear power plant, a probabilistic evaluation of loss-of-coolant accidents (LOCA) to identify the risks associated with them and to determine their acceptability in accordance with norms. This study includes all ruptures in the primary system that produce the automatic activation of 'emergency core cooling system'. Three starting events were selected for the probabilistic evaluation: 100% rupture of an input collector; 5% rupture of an input collector; 1.2% rupture of an input collector. At this stage the evaluation is focussed on the identification and quantization of the main failure sequences that follow a LOCA and lead to an uncontrolled reactor state or 'core meltdown'. The most important contribution to the core meltdown due to LOCA is the failure of supplies that are required for the emergency core cooling system. (Author)

Loss-of-coolant accident analysis of the Savannah River new production reactor design

International Nuclear Information System (INIS)

Maloney, K.J.; Pryor, R.J.

1990-11-01

This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1978-04-01

The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

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.

Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

Energy Technology Data Exchange (ETDEWEB)

Baek, Joo S.; Diamond, David

2016-12-06

A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in the analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.

Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

International Nuclear Information System (INIS)

Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

1989-01-01

The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

International Nuclear Information System (INIS)

1980-01-01

The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

Validation of advanced NSSS simulator model for loss-of-coolant accidents

Energy Technology Data Exchange (ETDEWEB)

Kao, S.P.; Chang, S.K.; Huang, H.C. [Nuclear Training Branch, Northeast Utilities, Waterford, CT (United States)

1995-09-01

The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

Loss of Coolant Accident Simulation for the Top-Slot break at Cold Leg Focusing on the Loop Seal Reformation under Long Term Cooling with the ATLAS

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Rok; Park, Yu Sun; Bae, Byoung Uhn; Choi, Nam Hyun; Kang, Kyoung Ho; Choi, Ki Yong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In the present paper, loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS, which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors (PWRs) of an advanced power reactor of 1400 MWe (APR1400). The simulation was focused on the loop seal reformation under long term cooling condition. During a certain class of Loss of Coolant Accident (LOCA) in a PWR like an advanced power reactor of 1400 MWe (APR1400), the steam volume in the reactor vessel upper plenum and/or upper head may continue expanding until steam blows liquid out of the intermediate leg (U-shaped pump suction cold leg), called loop seal clearing (LSC), opening a path for the steam to be relieved from the break. Prediction of the LSC phenomena is difficult because they are varies for many parameters, which are break location, type, size, etc. This LSC is the major factor that affects the coolant inventory in the small break LOCA (SBLOCA) or intermediate break LOCA (IBLOCA). There is an issue about the loop seal reformation that liquid refills intermediate leg and blocks the steam path after LSC. During the SBLOCA or IBLOCA, the Emergency Core Cooling System (ECCS) is operated. For long term of the top slot small or intermediate break at cold leg, the primary steam condensation by SG heat transfer or SIP, SIT water flooding (reverse flow to loop seal) make loop seal reformation possibly. The primary pressure increase at the top core region due to the steam release blockage by loop seal reformation. And then core level decreases and partial core uncover may occur. The loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS. The loop seal clearing and loop seal reformation were occurred repeatedly.

Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

Energy Technology Data Exchange (ETDEWEB)

Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

2007-02-15

Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

Investigation of break location effects on thermal-hydraulics during intermediate break loss-of-coolant accident experiments at ROSA-III

International Nuclear Information System (INIS)

Koizumi, Yasuo; Tasaka, Kanji

1986-01-01

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25 % main recirculation pump suction line break (MRPS-B) experiments, the 21 % single-ended jet pump drive line break (JPD-B) experiment and the 15 % main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests. In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop. (author)

Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

International Nuclear Information System (INIS)

Khan, L.A.; Bokhari, I.H.; Raza, S.H.

1993-12-01

Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs

Loss-of-coolant accident mitigation for the Advanced Neutron Source Reactor

International Nuclear Information System (INIS)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr.

1994-01-01

A RELAP5 Advanced Neutron Source Reactor system model has been developed for the conceptual design safety analysis. Three major regions modeled are the core, the heat exchanger loops, and letdown/pressurizing system. The model has been used to examine design alternatives for mitigation of loss-of-coolant accident (LOCA) transients. The safety margins to the flow excursion limit and critical heat flux are presented. The results show that the core can survive an instantaneous double-ended guillotine of the core outlet piping break (610 mm-diameter) provided a cavitating venturi is employed. RELAP5 calculations were also used to determine the effects of using a non-instantaneous break opening times. Both break opening time and break formation characteristics were included in these parametric calculations. Accumulator optimization studies were also performed which suggest that an optimum accumulator bubble size exists which improves system performance under some break scenarios

Loss of coolant analysis for the tower shielding reactor 2

International Nuclear Information System (INIS)

Radcliff, T.D.; Williams, P.T.

1990-06-01

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

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

Description of steam-condensation phenomena during the loss-of-coolant accident

International Nuclear Information System (INIS)

McCauley, E.W.; Holman, G.S.; Aust, E.; Schwan, H.; Vollbrandt, J.; Fuerst, H.

1980-01-01

The development and verification of advanced computer models which describe the boiling water reactor (BWR) pressure suppression process for a hypothetical loss-of-coolant accident (LOCA) require a clear description of basic steam condensation phenomena. The GKSS Research Center, in coordination with interested institutions of West Germany and the United States, is currently conducting a test program for such basic research on a multivent BWR-related pressure suppression system. The Lawrence Livermore National Laboratory (LLNL) acts as the principal US NRC liaison for this test program, with particular emphasis on development of GKSS data for confirmatory use regarding US Mark II nuclear power plants as well as to advanced code development. The multivent test facility, placed in operation in February 1979, is a three-pipe full-scale vent system modelling main features of both the West German KWU and United States G.E. Mk II BWR pressure suppression systems. The test facility and testing programs are described

Break spectrum analyses for small break loss of coolant accidents in a RESAR-3S Plant

International Nuclear Information System (INIS)

Fletcher, C.D.; Kullberg, C.M.

1986-03-01

A series of thermal-hydraulic analyses were performed to investigate phenomena occurring during small break loss-of-coolant-accident (LOCA) sequences in a RESAR-3S pressurized water reactor. The analysis included simulations of plant behavior using the TRAC-PF1 and RELAP5/MOD2 computer codes. Series of calculations were performed using both codes for different break sizes. The analyses presented here also served an audit function in that the results shown here were used by the US Nuclear Regulatory Commission (NRC) as an independent confirmation of similar analyses performed by Westinghouse Electric Company using another computer code. 10 refs., 62 figs., 14 tabs

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

International Nuclear Information System (INIS)

Kanzleiter, T.F.

1976-01-01

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1 : 64 compared to the 1200 MW PWR plant Biblis A. (Auth.)

Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

2000-01-01

Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 – Aluminum

Energy Technology Data Exchange (ETDEWEB)

Howe, Kerry J., E-mail: howe@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Mitchell, Lana, E-mail: lmitchell@alionscience.com [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kim, Seung-Jun, E-mail: skim@lanl.gov [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Blandford, Edward D., E-mail: edb@unm.edu [University of New Mexico, 210 University Blvd., Albuquerque, NM 87131 (United States); Kee, Ernest J., E-mail: erniekee@gmail.com [South Texas Project Nuclear Operating Company, P.O. Box 270, Wadsworth, TX 77483 (United States)

2015-10-15

Highlights: • Trisodium phosphate (TSP) causes aluminum corrosion to cease after 24 h of exposure. • Chloride, iron, and copper have a minimal effect on the rate of aluminum corrosion when TSP is present. • Zinc can reduce the rate of aluminum corrosion when TSP is present. • Aluminum occasionally precipitates at concentrations lower than the calculated solubility for Al(OH){sub 3}. • Corrosion and solubility equations can be used to calculate the solids generated during a LOCA. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum from metallic aluminum surfaces under conditions representative of the containment pool following a postulated loss of coolant accident at a nuclear power generating facility. The experiments showed that TSP is capable of passivating the aluminum surface and preventing continued corrosion after about 24 h at the conditions tested. A correlation that describes the rate of corrosion including the passivation effect was developed from the bench experiments and validated with a separate set of experiments from a different test system. The saturation concentration of aluminum was shown to be well described by the solubility of amorphous aluminum hydroxide for the majority of cases, but instances have been observed when aluminum precipitates at concentrations lower than the calculated aluminum hydroxide solubility. Based on the experimental data and previous literature, an equation was developed to calculate the saturation concentration of aluminum as a function of pH and temperature under conditions representative of a loss of coolant accident (LOCA) in a TSP-buffered pressurized water reactor (PWR) containment. The corrosion equation and precipitation equation can be used in concert with each other to calculate the quantity of solids that would form as a function of time during a LOCA if the temperature and pH profiles were known.

Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 1 – Aluminum

International Nuclear Information System (INIS)

Howe, Kerry J.; Mitchell, Lana; Kim, Seung-Jun; Blandford, Edward D.; Kee, Ernest J.

2015-01-01

Highlights: • Trisodium phosphate (TSP) causes aluminum corrosion to cease after 24 h of exposure. • Chloride, iron, and copper have a minimal effect on the rate of aluminum corrosion when TSP is present. • Zinc can reduce the rate of aluminum corrosion when TSP is present. • Aluminum occasionally precipitates at concentrations lower than the calculated solubility for Al(OH) 3 . • Corrosion and solubility equations can be used to calculate the solids generated during a LOCA. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum from metallic aluminum surfaces under conditions representative of the containment pool following a postulated loss of coolant accident at a nuclear power generating facility. The experiments showed that TSP is capable of passivating the aluminum surface and preventing continued corrosion after about 24 h at the conditions tested. A correlation that describes the rate of corrosion including the passivation effect was developed from the bench experiments and validated with a separate set of experiments from a different test system. The saturation concentration of aluminum was shown to be well described by the solubility of amorphous aluminum hydroxide for the majority of cases, but instances have been observed when aluminum precipitates at concentrations lower than the calculated aluminum hydroxide solubility. Based on the experimental data and previous literature, an equation was developed to calculate the saturation concentration of aluminum as a function of pH and temperature under conditions representative of a loss of coolant accident (LOCA) in a TSP-buffered pressurized water reactor (PWR) containment. The corrosion equation and precipitation equation can be used in concert with each other to calculate the quantity of solids that would form as a function of time during a LOCA if the temperature and pH profiles were known

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)

Investigation of small break loss-of-coolant phenomena in a small scale nonnuclear test facility

International Nuclear Information System (INIS)

Cozzuol, J.M.; Fauble, T.J.; Harvego, E.A.

1980-01-01

A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and included initial conditions which were similar to conditions in a PWR operating at full power. The 2.5% break size ensured that the nominal break flow rate was greater than the high pressure injection system (HPIS) flow rate, thus providing the potential for a continuous system depressurization. The sequence of events was similar to that used in evaluation model analysis of small break loss-of-coolant accidents, and included simulated reactor scram and loss of offsite power. Comparisions of experimental data with computer code calculations are used to demonstrate the capabilities and limitations of integral system calculations used to predict phenomena which can be important in the assessment of a small break LOCA in a PWR

CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

International Nuclear Information System (INIS)

Haste, T.J.

1982-07-01

The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

International Nuclear Information System (INIS)

Megahed, M.M.

1992-01-01

An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

Theoretical analysis of the temperature changes and resultant loss of fuel integrity in the IEA-R1 research reactor fuel elements following a loss of coalant accident

International Nuclear Information System (INIS)

Garone, J.G.M.

1983-01-01

The IEA-R1 core following a loss of coolant accident (LOCA) is analysed. THe AIRLOCA code was used to calculate fuel temperatures, heat generation due to fission product decay and convective and radiative heat transfer from the fuel elements to the surrounding air both during and following the loss of coolant. The influence of certain critical parameters, such as log time, specific power was studied in detail. Representative results are presented and suggestions made to ensure that fuel integrity is maintained following a LOCA. (Author) [pt

Analysis of the effects of the pressure wave generated in loss of coolant accidents in reactor vessels

International Nuclear Information System (INIS)

Valero Martinez, M.

1980-01-01

The increasing demands in the field of ''Nuclear Safety'', obliges to a perfect knowledge of the causes and effects of every possible accident in a nuclear power plant. In this paper will be analysed the effects of the pressure wave appearing in a LOCA (Loss of collant accident). The pressure wave could deform the following structures: core barrel wall, cover and bottom, control rods and safety coolant system. Any change of the geometry of these structures could provoke and incorrect system reaction after the accident has happened. The basis and hypothesis for the theoretical analysis will be exposed. The structures are considered to be rigid. A typical boiling water be analysed and the developed theory will be verified in comparations with experimental results and the results obtained with some others models. Due to the easy application and short calculation time of the created programmes, they are recommended for parametrical calculations in the analysis of the pressurized water reactors and boiling water reactors. (author)

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

Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

International Nuclear Information System (INIS)

Rivas, J.C.; Dies, J.; Fajarnés, X.

2015-01-01

Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

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

Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

International Nuclear Information System (INIS)

Pitts, J.H.; McCauley, E.W.

1977-01-01

A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

Generic evaluation of feedwater transients and small break loss-of-coolant accidents in GE-designed operating plants and near-term operating license applications

International Nuclear Information System (INIS)

1980-01-01

The results are presented of a generic evaluation of feedwater transients, small-break loss-of-coolant accidents (LOCAs), and other TMI-2-related events for General Electric Company (GE)-designed operating plants and near-term operating license applications to confirm or establish the bases for the continued safe operation of the operating plants. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas. Additional review of the accident is continuing and further information is being obtained and evaluated. Any new information will be reviewed and modifications will be made as appropriate

Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Suman, Siddharth [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Khan, Mohd Kaleem, E-mail: mkkhan@iitp.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Pathak, Manabendra [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Singh, R.N.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

2016-10-15

Highlights: • Modelling of nuclear fuel cladding during loss-of-coolant accident transient. • Phase transformation, corrosion, and creep combined to evaluate burst criterion. • Effect of oxygen concentration on burst stress and burst strain. • Effect of heating rate, internal pressure fluctuation, shear modulus incorporated. - Abstract: A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

A review of Zircaloy fuel cladding behavior in a loss-of-coolant accident

International Nuclear Information System (INIS)

Erbacher, F.J.; Leistikow, S.

1985-09-01

The paper reviews the state-of-the-art experimental work performed in several countries with respect to the acceptance criteria established for emergency core cooling (ECC) in a loss-of-coolant accident (LOGA) of light water reactors (LWRs). It covers in detail oxidation, embrittlement, plastic deformation and coolability of deformed rod bundles. The main test results are discussed on the basis of research work performed at the Karlsruhe Nuclear Research Center (KfK) within the framework of the Nuclear Safety Project (PNS) and reference is made to test data obtained in other countries. The conclusion reached in the paper is that the major mechanisms and consequences of oxidation, deformation and emergency core cooling are sufficiently investigated in order to provide a reliable data base for safety assessments and licensing of LWRs. All test data prove that the ECC-criteria are conservative and that the coolability of an LWR and the public safety can be maintained in a LOCA. (orig.) [de

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

Energy Technology Data Exchange (ETDEWEB)

Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

2014-12-15

Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

International Nuclear Information System (INIS)

Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

2014-01-01

Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

Prediction of thermal hydraulic parameters in the loss of coolant accident by using artificial neural networks

International Nuclear Information System (INIS)

Vaziri, N.; Erfani, A.; Monsefi, M.; Hajabri, A.

2008-01-01

In a reactor accident like loss of coolant accident , one or more signals may not be monitored by control panel for some reasons such as interruptions and so on. Therefore a fast alternative method could guarantee the safe and reliable exploration of nuclear power planets. In this study, we used artificial neural network with Elman recurrent structure to predict six thermal hydraulic signals in a loss of coolant accident after upper plenum break. In the prediction procedure, a few previous samples are fed to the artificial neural network and the output value or next time step is estimated by the network output. The Elman recurrent network is trained with the data obtained from the benchmark simulation of loss of coolant accident in VVER. The results reveal that the predicted values follow the real trends well and artificial neural network can be used as a fast alternative prediction tool in loss of coolant accident

Analysis of the behaviour of pressure and temperature of the containment of a PWR reactor, submitted to a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Silva, D.E. da; Arrieta, L.A.J.; Costa, J.R.; Camargo, C.; Santos, C.M. dos; Rochedo, E.R.R.

1979-12-01

The main purpose of this work is to analyse the pressure and temperature behaviour of the metalic containment of a PWR building, submitted to a postulated loss-of-coolant accident (LOCA) caused by a double-ended rupture in the main line of the primary circuit. The scope of the study was directed to verify the Final Safety Analysis Report (FSAR) results for the integrity of the metalic containment of the Angra I power plant. The highest containment pressure peak for this unit is expected for a break in the suction line of one of the main pumps of the primary coolant. Using the same input data, our results are very similar to those presented in the FSAR which shows a reasonable equivalence between the two analytical models. Using as input data the results of a previous LOCA study at CNEN, which yields to more conservative boundary conditions than those presented by the FSAR, the pressure and temperature peak values determined by our model are quite larger than those presented by the cited Safety Report. (author) [pt

The condensation of steam on the external surfaces of the shells of HIFAR heavy water heat exchangers during a loss-of-coolant accident

International Nuclear Information System (INIS)

Chapman, A.G.

1987-03-01

A study of steam condensation rates on the HIFAR heavy water heat exchangers was undertaken to predict thermohydraulic conditions in the HIFAR containment during a postulated loss-of-coolant accident (LOCA). The process of surface condensation from a mixture of air and steam, and methods for calculating the rate of condensation, are briefly reviewed. Suitable experimental data are used to estimate coefficients of condensation heat transfer to cool surfaces in a reactor containment during a LOCA. The relevance of the available data to a LOCA in the HIFAR materials testing reactor is examined, and two sets of data are compared. The differences between air/H 2 O and air/D 2 O mixtures are discussed. Formulae are derived for the estimation of the coefficient of heat transfer from the heat exchanger shells to the cooling water, and a method of calculating the rate of condensation per unit area of surface is developed

An on-line pressurizer surveillance system design to prevent small-break loss-of-coolant accidents through power-operated relief valves using a microcomputer

International Nuclear Information System (INIS)

Lee, J.H.; Chang, S.H.

1987-01-01

A small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve is one of the important contributors to nuclear power plant risk. A pressurizer surveillance system was designed to use a microcomputer to prevent the malfunction of the system; the effect of this improvement has been assessed through probabilistic risk assessment. The microcomputer diagnoses the malfunction of the system by a process-checking method and automatically performs the backup action related to each malfunction. This improvement means that we can correctly diagnose ''spurious opening,'' ''failure to reclose,'' and ''small-break LOCA,'' which are difficult for operators to diagnose quickly and correctly, and by taking automatic backup action one can reduce the probability of human error

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

Analysis of risk reduction methods for interfacing system LOCAs [loss-of-coolant accidents] at PWRs

International Nuclear Information System (INIS)

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

1988-01-01

The Reactor Safety Study (WASH-1400) predicted that Interfacing System Loss-of-Coolant Accidents (ISL) events were significant contributors to risk even though they were calculated to be relatively low frequency events. However, there are substantial uncertainties involved in determining the probability and consequences of the ISL sequences. For example, the assumed valve failure modes, common cause contributions and the location of the break/leak are all uncertain and can significantly influence the predicted risk from ISL events. In order to provide more realistic estimates for the core damage frequencies (CDFs) and a reduction in the magnitude of the uncertainties, a reexamination of ISL scenarios at PWRs has been performed by Brookhaven National Laboratory. The objective of this study was to investigate the vulnerability of pressurized water reactor designs to ISLs and identify any improvements that could significantly reduce the frequency/risk of these events

Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

International Nuclear Information System (INIS)

Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

2010-01-01

For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

Loss of coolant accident. Past, present and future

International Nuclear Information System (INIS)

Cermak, J.O.

1978-01-01

The history of LOCA is covered from the original design basis, failure of the largest connecting pipe with no peak clad temperature limit through to the current design basis, double ended failure of the largest pipe in the primary system with a design peak clad temperature limit of 2200 0 F. Various obstacles along the way are addressed such as, degree of analytical sophisticaton, perplexing experimental results, the infamous 1971 semiscale tests, fuel densification and changes in USNRC evaluation models. In the future, it is projected that more reliance will be put on probabilistic evaluation of the LOCA with respect to both the accident analysis, reliability of the system and the probability of the accident itself. (author)

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

Review on research of small break loss of coolant accident

International Nuclear Information System (INIS)

Bo Jinhai; Wang Fei

1998-01-01

The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

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)

Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Conlon, W.M.; Lahey, R.T. Jr.

1982-12-01

This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

Microstructural examination of fuel rods subjected to a simulated large-break loss of coolant accident in reactor

International Nuclear Information System (INIS)

Garlick, A.

1985-01-01

A series of tests has been conducted in the National Research Universal (NRU) reactor, Chalk River, Canada, to investigate the behaviour of full-length 32-rod PWR fuel bundles during a simulated large-break loss of coolant accident (LOCA). In one of these tests (MT-3), 12 central rods were pre-pressurized in order to evaluate the ballooning and rupture of cladding in the Zircaloy high-α/α+β temperature region. All 12 rods ruptured after experiencing < 90% diametral strain but there was no suggestion of coplanar blockage. Post-irradiation examination was carried out on cross-sections of cladding from selected rods to determine the aximuthal distribution of wall thinning along the ballooned regions. These data are assessed to check whether they are consistent with a mechanism in which fuel stack eccentricity generates temperature gradients around the ballooning cladding and leads to premature rupture during a LOCA. After anodizing, the cladding microstructures were examined for the presence of prior-beta phase that would indicate the α/α+β transformation temperature (1078K) had been exceeded. These results were compared with isothermal annealing test data on unirradiated cladding from the same manufacturing batch

Responses to Small Break Loss of Coolant Accidents for SMART

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

2004-01-01

The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

Nonstationary pressure build up in full-pressure containments after a loss-of-coolant accident

International Nuclear Information System (INIS)

Mansfeld, G.

1977-01-01

The time histories of pressure, temperature and pressure difference during the pressure build up phase of a loss-of-coolant accident (LOCA) in the primary system in full-pressure containments of water cooled nuclear power reactors are treated. These are important for the design of such containments. The experiments within the German research program RS 50 ''Druckverteilung im Containment'' offered, for the first time, the opportunity to observe experimentally fluid-dynamic processes in a multiple divided full-pressure containment, and to test at the same time, computer codes which serve to describe the physical processes during the LOCA. The comparison of the results calculated by the computer codes ZOCO VI and DDIFF with the experimental results showed apparent deviations by special arrangements of the compartments and the vent flow paths of a model containment for the calculation of time dependent pressure-, temperature- and pressure difference-histories. The deviations lead to the development of the analytical model and computer code COFLOW. This new model was primarily designed to deal with the fluid-dynamic processes in the beginning phase of the blowdown as maximal pressure differences appear. Furthermore, it can be used to determine the maximum containment pressure, as well as for long term calculations. The analytical model and computer code COFLOW shows a better correlation between theory and experiment than previous codes

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

Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

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

1986-09-01

The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

International Nuclear Information System (INIS)

Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-01-01

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-04-15

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

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

Thermal hydraulic analysis of aggressive secondary cooldown in small break loss of coolant accident with total loss of high pressure safety injection

International Nuclear Information System (INIS)

Han, S. J.; Im, H. K.; Yang, J. U.

2003-01-01

Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). To use RIA, the present study focuses on the detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study is to evaluate the success criteria of Aggressive Secondary Cooldown (ASC) in Small Break Loss Of Coolant Accident (SBLOCA) with total loss of High Pressure Safety Injection (HPSI) and to enhance the understanding of related thermal hydraulic behavior and phenomena. The accident scenario was 2 inch coldleg break LOCA without HPSI, with 1/2 Low Pressure Safety Injection (LPSI), and performing ASC limited by 55.6 .deg. C /hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip, which successively reaches the LPSI condition for about 1.5hr after starting ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria 1204.4 .deg. C (2200 .deg. F). In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that operator should maintain the adequate ASC operation. However, it is necessary to evaluate uncertainties arisen from the related parameters of the ASC operation

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)

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

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

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)

Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

International Nuclear Information System (INIS)

Luo Bangqi

2012-01-01

The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

Analysis of LOFT loss-of-coolant experiments L2-2, L2-3, and L3-0

International Nuclear Information System (INIS)

Leach, L.P.; Linebarger, J.H.

1979-01-01

A summary of results from Loss-of-Coolant Experiments (LOCE) L2-2, L2-3, and L3-0, conducted in the Loss-of-Fluid Test (LOFT) facility, and conclusions from posttest analyses of the experimental data are presented. LOCEs L2-2 and L2-3 were nuclear large break experiments and were dominated by a core-wide fuel rod cladding rewet, which limited the maximum fuel temperature. Analytical models only conservatively predicted the measured fuel rod temperatures and will require improvements to provide best estimate predictions in this area. Analysis of a large commercial pressurized water reactor (PWR) indicates that the cladding rewet observed in LOFT is also likely to occur in a large PWR, and that, therefore, safety analysis calculations of large loss-of-coolant accidents (LOCA) are more conservative than previously thought. LOCE L3-0 was an isothermal small break (top of pressurizer) experiment and illustrated that the pressurizer fills after the primary system fluid saturates someplace other than the pressurizer itself, that the indicated pressurizer level is higher than the actual level, and that additional model development and assessment work is necessary in order to predict small LOCAs as accurately as large LOCAs

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

RELAP5 simulation of a large break Loss of Coolant Accident (LOCA) in the hot leg of the primary system in Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Andrade, Delvonei Alves de; Sabundjian, Gaiane

2004-01-01

The objective of this work is to present the simulation of a large break loss of coolant accident - LBLOCA in the hot leg of the primary loop in Angra 2, with RELAP5/MOD3.2.2g code. This accident is described in the Final Safety Report Analysis of Angra 2 - FSAR and consists basically of the hot leg total break, in loop 20 of the plant. The area considered for the rupture is 4480 cm 2 , which corresponds to 100% of the pipe flow area. Besides, this work also has the objective of verifying the efficiency of the emergency core coolant system - ECCS in case of accidents and transients. The thermal-hydraulic processes inherent to the accident phenomenology, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the liquid level, until the ECCS is capable to reflood it

Loss of coolant accident analysis (thermal hydraulic analysis) - Japanese industries experience

International Nuclear Information System (INIS)

Okabe, K.

1995-01-01

An overview of LOCA analysis in Japanese industry is presented. The BASH-M code, developed for large scale LOCA reflooding analysis, is given as an example of verification and improvement of US computer programs are given. The code's application to the operational safety analysis concerns the following main areas: 1D drift flux model base computer program CANAC; CANAC-based advanced training simulator; emergency operating procedures. The author considers also the code application to the following new PWR safety design concepts: use of steam generators for decay heat removal at LOCA conditions; use of horizontal type steam generator for maintaining two-phase natural circulation under the reactor coolant system submerged. 9 figs

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)

Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

Directory of Open Access Journals (Sweden)

Omid Noori-Kalkhoran

2016-10-01

Full Text Available Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model. In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

International Nuclear Information System (INIS)

Kalkahoran, Omid Noori; Ahangari, Rohollah; Shirani, Amir Saied

2016-01-01

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results

Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

Energy Technology Data Exchange (ETDEWEB)

Kalkahoran, Omid Noori; Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

2016-10-15

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

International Nuclear Information System (INIS)

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

1986-09-01

The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

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.)

Loss of coolant accident at boiling water reactors

International Nuclear Information System (INIS)

Ramirez G, R.

1975-01-01

A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

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)

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

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)

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)

Accident beyond the design basis management with the coolant loss at the NPP with WWER

International Nuclear Information System (INIS)

Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

2010-01-01

The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

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)

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

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

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)

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.

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

Environmental radiological consequences of a loss of coolant accident

International Nuclear Information System (INIS)

Guimaraes, A.C.F.

1981-01-01

The elaboration of a calculation model to determine safety areas, named Exclusion Zone and Low Population Zone for nuclear power plants, is dealt with. These areas are determined from a radioactive doses calculation for the population living around the NPP after occurence of a postulated ' Maximum Credible Accident' (MCA). The MCA is defined as an accident with complete loss of primary coolant and consequent fusion of a substantial portion of the reactor core. In the calculations carried out, data from NPP Angra I were used and the assumptions made were conservative, to be compatible with licensing requirements. Under the most pessimistic assumption (no filters) the values of 410m and 1000m were obtained for the Exclusion Zone and Low Population Zone radii, respectivily. (Author) [pt

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

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

Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

International Nuclear Information System (INIS)

Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

2003-03-01

Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). The present study focuses on detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model using RIA for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study in this year is to evaluate the success cri-teria of Aggressive Secondary Cooldown (ASC) in a Small Size Loss Of Coolant Accident (SBLOCA) without HPSI and to enhance the understanding of related thermal hydraulic behavior and phenomena. An effort was made to evaluate the system success criteria and a mission time for the recovery action by an operator to prevent the core damage for that accident scenario. The accident scenario for KSNP was a 2 inch coldleg break LOCA with a total loss of High Pressure Safety Injection (HPSI) and 1/2 Low Pressure Safety Injection (LPSI) available and perform-ing ASC limited by 55.6 .deg. C/hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip. It successively reached the LPSI condition for about 1.5hr after starting the ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria of 1204.4 .deg. C (2200 .deg. F). Sensitivity studies were performed for (1) cool-ant average temperature parameters, (2) ASC operation control method, (3) operation start time, (4) 1 inch break size. The present analysis identified thermal hydraulic phenomena and parameters affecting on the behavior, which consist of coolant break flow and inventory, parameters governing secondary heat removal, ASC operation control method, and its reference temperature parameters. In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that an operator should maintain the ade-quate ASC operation. However, it is necessary to evaluate the uncertainties arisen from the

Analysis of a simulated small break in the semiscale system under loss-of-coolant accident conditions

International Nuclear Information System (INIS)

Cartmill, C.E.

1978-01-01

The Semiscale Mod-1 experimental program conducted by EG and G Idaho, Inc., is part of the overall U.S. Nuclear Regulatory Commission (NRC) and Department of Energy (DOE) sponsored research and development program to investigate the behavior of the pressurized water reactor (PWR) system during an hypothesized loss-of-coolant accident (LOCA). The Semiscale Mod-1 program is intended to provide transient thermal-hydraulic data from a simulated LOCA using a small-scale experimental nonnuclear system. The Semiscale Mod-1 program is a major contributor of experimental data that provide a means of evaluating the adequacy of overall system analytical models as well as the models of the individual system components. Selected experimental data produced by this program will also be used to aid other DOE and NRC sponsored experimental programs, such as the Loss-of-Fluid Test (LOFT) program in optimizing test series, selecting test parameters, and evaluating test results. The Semiscale Mod-1 tests are performed with an experimental system which simulates the principal features of a nuclear plant but which is smaller in volume. Nuclear heating is simulated in the tests by a core composed of an array of electrically heated rods. The core is contained in a pressure vessel which also includes a downcomer, lower plenum, and upper plenum. The Semiscale system piping is arranged such that the intact loop represents three loops of a four-loop nuclear plant, and the broken loop represents the fourth loop. In the present configuration the intact loop contains an active steam generator and pump, and the broken loop contains passive simulators for the steam generator and pump

Confinement barriers for loss of coolant accidents in the SEAFP reactor plant models

International Nuclear Information System (INIS)

Blomquist, R.; Ebert, E.; Gay, J.M.; Mazille, F.; Natalizio, A.; Rolandsson, S.; Ross, W.E.; Shen, K.; Sjoeberg, A.

1995-01-01

Loss of coolant accidents may mobilise radioactivity and pressurise confinement barriers thereby making a release to the environment possible. The paper defines the radioactivity confinements and presents principal results from the underlying thermal-hydraulic analyses. (orig.)

Strategies for reactor safety: Preventing loss of coolant accidents. Final report

International Nuclear Information System (INIS)

Lydell, B.O.Y.

1997-12-01

This final report on the NKS/RAK-1.2 summarizes the main features of the PIFRAP PC-program and its intended implementation. Regardless of the preferred technical approach to LOCA frequency estimation, the analysis approach must include recognition of the following technical issues: a) Degradation and failure mechanisms potentially affecting piping systems within the reactor coolant pressure boundary (RCPB) and the potential consequences; b) In-service inspection practices and how they influence piping reliability; and c) The service experience with piping systems. The report consists of six sections and one appendix. A Nordic perspective on LOCA and nuclear safety is given. It includes summaries of results from research in material sciences and current regulatory philosophies regarding piping reliability. A summary of the LOCA concept is applied in Nordic PSA studies. It includes a discussion on deterministic and probabilistic views on LOCA. The R and D on piping reliability by SKI and the PIFRAP model is summarized. Next, Section 6 presents conclusion and recommendations. Finally, Appendix A contains a list of abbreviations and acronyms, together with a glossary of technical terms. (EG)

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)

Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

A 200-percent double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops (three intact, one broken) and cold-leg emergency-core-cooling systems (ECCS). The finely noded TRAC model employed 440 three dimensional (r, THETA, z) vessel cells along with approximately 300 one-dimensional cells that modeled the primary system loops. The calculated peak-clad temperature of 950 0 K occurred during blowdown and the clad temperature excursion was terminated at 175 s, when complete core quenching occurred. Accumulator flows were initiated at 10 s, when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated

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

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.

Accident sequences simulated at the Juragua nuclear power plant

International Nuclear Information System (INIS)

Carbajo, J.J.

1998-01-01

Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident (LOCA) with the emergency core coolant system (ECCS) on, (2) a station blackout (SBO), (3) a small LOCA (SLOCA) concurrent with SBO, (4) a large LOCA (LLOCA) concurrent with SBO, and (5) a LLOCA concurrent with SBO and with the containment breached at time zero. Timings of important events and source term releases have been calculated for the different sequences analyzed. Under certain weather conditions, the fission products released from the severe accident sequences may travel to southern Florida

Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

2000-01-01

The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

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

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

CONTEMPT-LT/028: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hargroves, D.W.; Metcalfe, L.J.; Wheat, L.L.; Niederauer, G.F.; Obenchain, C.F.

1979-03-01

CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program 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 to describe fan cooler and cooling spray engineered safety systems. An annular fan model is also provided to model pressure control in the annular region of dual containment systems. Up to four compartments can be modeled with CONTEMPT-LT, 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

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

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

Study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS using RELAP5 code

International Nuclear Information System (INIS)

Hoang Minh Giang; Ha Thi Anh Dao; Hoang Tan Hung; Bui Thi Hoa; Nguyen Thi Tu Oanh; Dinh Anh Tuan; Pham Tuan Nam

2017-01-01

The advanced VVER-1200/V491 reactor designed with passive safety systems to deal with design extension conditions is primarily selected as priority candidate for Ninh Thuan 1 nuclear power plant project. So that, in order to enhance competence of nuclear safety and toward participation on review Safety Analysis Report (SAR) of Ninh Thuan nuclear Power project the study on safety analysis of VVER-1200/V491 in scenario of Loss of Coolant Accidents along with partly failure of ECCS is implemented. As requirement of the study, the input deck file of VVER-1200/V491 for RELAP5 and analysis report for some special case of LOCAs along with partly failure of ECCS are issued. (author)

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.

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

A study of entrainment at a break and in the core during small break loss-of-coolant accidents in PWRs

International Nuclear Information System (INIS)

Yonomoto, Taisuke

1996-05-01

Objectives of the present study are to obtain a better understanding of entrainment at a break and in the core during small break loss-of-coolant-accidents (SBLOCAs) in PWRs, and to develop a means for the best evaluation of the phenomena. For the study of entrainment at a break, a theoretical model was developed, which was assessed by comparisons with several experimental data bases. By modifying a LOCA analysis code using the present model, experimental results obtained from SBLOCA experiments at a PWR large-scale simulator were reproduced very well. For the study of entrainment in the core, reflooding experiments were conducted at high pressure, from which the onset conditions were obtained. It was confirmed that the cooling behavior for a dry-out core is very simple under typical high pressure reflooding conditions for PWRs, because liquid entrainment does not occur in the core. (author)

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.

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.

Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

2015-07-01

The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

Theoretical study on loss of coolant accident of a research reactor

International Nuclear Information System (INIS)

Lee, Kwon-Yeong; Kim, Wan-Soo

2016-01-01

Highlights: • A theoretical model of siphon breaking phenomena was developed. • A general formula using Chisholm coefficient B was proposed. • The safety requirements regarding a loss of coolant accident of research reactors could be found out. - Abstract: Under the design conditions of a research reactor, the siphon phenomenon induced by pipe rupture can cause continuous efflux of water. In order to prevent water efflux, an additional facility is necessary. A siphon breaker is a type of safety facility that can resist the loss of coolant effectively. However, analysis of siphon breaking is complex since it comprises two-phase flow and there are many inputs to be considered. For this reason, we analyzed the experimental results to develop a theoretical model of siphon breaking phenomena. Developed model is based on fluid mechanics and Chisholm model. From Bernoulli’s equation, the velocity and quantity as well as undershooting height, water level, pressure, friction coefficient, and factors related to the two-phase flow could be calculated. The Chisholm model, which is able to analyze the two-phase flow, can predict the results in a manner similar to those obtained from a real-scale experiment, and a general formula using Chisholm coefficient B was proposed in this study. Also, we verified the theoretical model and concluded that it is possible to analyze the siphon breaking. Moreover, the design conditions that can satisfy the safety requirements regarding a loss of coolant accident of research reactors could be found out by using the theoretical model. In conclusion, we propose the theoretical model which can analyze the siphon breaking as real, and it is helpful not only to analyze but also to design the siphon breaker.

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

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

A study of the loss of coolant accident

International Nuclear Information System (INIS)

Lee, Y.W.; Chung, M.K.; Kim, S.H.; Park, J.S.; Lee, C.B.; Kim, S.B.; Won, S.Y.; Cho, Y.R.

1983-01-01

The primary objectives of this project are: (1) To review the published information on LOCA/ECCS study (2) To investigate reflood phenomena and to provide necessary information for analytical model development (3) To modyfy and develop a reflood analysis code. To review the published information on LOCA/ECCS, heat transfer phenomena are divided into 4 regions. Heat transfer correlations published in the references are reviewed and classified according to the regions. To investigate reflood phenomena and to provide better modeling of reflood phenomena, experments have been carried out with an electrically heated 3x3 rod bundle. Heat flux and heat transfer coefficients at the hot surface have been determined from the experimental data by HTC program. The influences of the parameters such as flooding rate, coolant subcooling and power generation on the propagation of rewetting front were also investigated. Calculations obtained from REFLUX code were compared with the experimental data to help an understanding of the reflood heat transfer mechanisms, and then some modifications of the code were provided. Improvements in heat transfer correlations of transition and inverted annular film boiling region, and the logic for the selection of heat transfer regime allowed better estimate for rod temperature behavior. (Author)

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

Description of steam condensation phenomena during the loss-of-coolant accident

International Nuclear Information System (INIS)

McCauley, E.W.; Holman, G.S.; Aust, E.; Furst, H.; Schwan, H.; Vollbrandt, J.

1981-01-01

Study of results from the full scale multivent pressure suppression experiment conducted by the GKSS Laboratory has developed an improved understanding of the dynamic, oscillatory steam condensation events and related loading functions which occur during the hypothetical loss-of-coolant accident in a boiling water nuclear reactor. Due to the unique measurements systems which combines both cinematic and digital data, qualified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena has been obtained

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

Severe accident sequences simulated at the Grand Gulf Nuclear Station

International Nuclear Information System (INIS)

Carbajo, J.J.

1999-01-01

Different severe accident sequences employing the MELCOR code, version 1.8.4 QK, have been simulated at the Grand Gulf Nuclear Station (Grand Gulf). The postulated severe accidents simulated are two low-pressure, short-term, station blackouts; two unmitigated small-break (SB) loss-of-coolant accidents (LOCAs) (SBLOCAs); and one unmitigated large LOCA (LLOCA). The purpose of this study was to calculate best-estimate timings of events and source terms for a wide range of severe accidents and to compare the plant response to these accidents

Radiological impact of a loss of coolant accident at Angra 2 reactor

International Nuclear Information System (INIS)

Dias, W.

1992-01-01

A loss of coolant accident is analyzed which comprises a double ended rupture of a main primary system line. The accident sequence is described and the main assumptions as to the activity release are presented. On the basis of site specific meteorological data, the atmospheric dispersion factors are calculated using the Gaussian plume diffusion model and the doses are then determined at the boundary of the low population zone. The resulting values for the effective dose equivalent are more than one order of magnitude below that due to the average background radiation received in one year. (author)

Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

International Nuclear Information System (INIS)

Adam, E.; Carl, H.; Kubis, K.

1979-01-01

Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

International Nuclear Information System (INIS)

Austregesilo Filho, H.

1978-12-01

The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author) [pt

Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

International Nuclear Information System (INIS)

Zhang Longfei; Zhang Dafa; Wang Shaoming

2007-01-01

Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

Evaluation of a postulated loss of coolant accident (LOCA) due to a 160 cm2 break in a cold leg of Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Azevedo, Carlos Vicente Goulart de; Palmieri, Elcio Tadeu; Aronne, Ivan Dionysio

2002-01-01

The development of a qualified full nodalization of Angra2 NPP for RELAP5/Mod 3.2.2 gamma, aiming at the evaluation of a comprehensive number of accidents and transients, thus providing suitable safety analysis support for licensing purposes, is being carried out within the framework of CNEN internal technical cooperation, involving some of its institutes (CDTN, IPEN and IEN) and the Reactors Coordination (CODRE). This work presents a simulation of a postulated Angra2 small cold leg break loss of coolant accident (SBLOCA). A 160 cm 2 break is supposed to occur at one cold leg between the main coolant pump and the reactor vessel and is described in the Angra2 Final Safety Analysis Report, section 15.6.4.1.3.4. The simulation of several types of transients and accidents is necessary to verify the adequate performance of the modeled logic and systems. In general, the analysis of such and accident allows to demonstrate the safety Injection System performance and the reliable transition between the high pressure safety injection, the accumulator injection and the residual heat removal phases. Furthermore, it is assumed that some components are out of service due to fail or repair in order to make a conservative analysis. The results showed a compatible behavior of the molded systems and that the simulated Emergency Core Cooling System was able to provide sufficient cooling to avoid any damage to the core. (author)

Nonlinear dynamic analysis of nuclear reactor primary coolant systems

International Nuclear Information System (INIS)

Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

1979-01-01

The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

International Nuclear Information System (INIS)

Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

2007-01-01

As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

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

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

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

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.)

A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident

International Nuclear Information System (INIS)

Gaballah, I.

1978-09-01

A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident. A theory was developed for the calculation of a dispersed two phase flow with heat addition in a channel with general area change. The theory was used to study different thermodynamic and gasdynamic processes, which may occur during the emergency cooling after a LOCA of a pressurized water reactor. The basic equations were formulated and solved numerically. The heat transfer mechanism was examined. Calculations have indicated that the radiative heat flux component is small compared to the convective component. A drop size spectrum was used in the calculations. Its effect on the heat transfer was investigated. It was found that the calculation with a mean drop diameter gives good results. Significant thermal non-equilibrium has been evaluated. The effect of different operating parameters on the degree of thermal non-equilibrium was studied. The flow and heat transfer in a channel with cross-sectional area change were calculated. It was shown that the channel deformation affects the state properties and the heat transfer along the channel very strongly. (orig.) 891 GL [de

Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 4 – Integrated chemical effects testing

Energy Technology Data Exchange (ETDEWEB)

Ali, Amir; LaBrier, Daniel [Department of Nuclear Engineering, University of New Mexico (United States); Blandford, Edward, E-mail: edb@unm.edu [Department of Nuclear Engineering, University of New Mexico (United States); Howe, Kerry [Department of Civil Engineering, University of New Mexico (United States)

2016-04-15

Highlights: • Integrated test explored the material release of a postulated large break LOCA. • Aluminum concentration was very low (<0.1 mg/L) throughout the test duration. • Zinc concentration was low (<1 mg/L) in TSP-buffered system. • Calcium release showed two distinguished release zones: prompt and meta-stable. • Copper and iron has no distinguishable concentration up to first 24 h of testing. - Abstract: This paper presents the results of an integrated chemical effects experiment executed under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at the Vogtle nuclear power plant, operated by the Southern Nuclear Operating Company (SNOC). This test was conducted for closure of a series of bench scale experiments conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum (Howe et al., 2015) and zinc (Pease et al., 2015) from metallic surfaces, and calcium from NUKON fiberglass insulation (Olson et al., 2015) . The integrated test was performed in the Corrosion/Chemical Head Loss Experimental (CHLE) facility with representative amounts of zinc, aluminum, carbon steel, copper, NUKON fiberglass, and latent debris. The test was conducted using borated TSP-buffered solution under a post-LOCA prototypical temperature profile lasting for 30 days. The results presented in this article demonstrate trends for zinc, aluminum, and calcium release that are consistent with separate bench scale testing and previous integrated tests under TSP conditions. The release rate and maximum concentrations of the released materials were slightly different than the separate effect testing as a result of different experimental conditions (temperature, surface area-to-water volume ratio) and/or the presence of other metals and chemicals in the integrated test. Samples of metal coupons and fiberglass were selected for analysis using Scanning Electron Microscopy

Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 4 – Integrated chemical effects testing

International Nuclear Information System (INIS)

Ali, Amir; LaBrier, Daniel; Blandford, Edward; Howe, Kerry

2016-01-01

Highlights: • Integrated test explored the material release of a postulated large break LOCA. • Aluminum concentration was very low (<0.1 mg/L) throughout the test duration. • Zinc concentration was low (<1 mg/L) in TSP-buffered system. • Calcium release showed two distinguished release zones: prompt and meta-stable. • Copper and iron has no distinguishable concentration up to first 24 h of testing. - Abstract: This paper presents the results of an integrated chemical effects experiment executed under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at the Vogtle nuclear power plant, operated by the Southern Nuclear Operating Company (SNOC). This test was conducted for closure of a series of bench scale experiments conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of aluminum (Howe et al., 2015) and zinc (Pease et al., 2015) from metallic surfaces, and calcium from NUKON fiberglass insulation (Olson et al., 2015) . The integrated test was performed in the Corrosion/Chemical Head Loss Experimental (CHLE) facility with representative amounts of zinc, aluminum, carbon steel, copper, NUKON fiberglass, and latent debris. The test was conducted using borated TSP-buffered solution under a post-LOCA prototypical temperature profile lasting for 30 days. The results presented in this article demonstrate trends for zinc, aluminum, and calcium release that are consistent with separate bench scale testing and previous integrated tests under TSP conditions. The release rate and maximum concentrations of the released materials were slightly different than the separate effect testing as a result of different experimental conditions (temperature, surface area-to-water volume ratio) and/or the presence of other metals and chemicals in the integrated test. Samples of metal coupons and fiberglass were selected for analysis using Scanning Electron Microscopy

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.

Comparison of the Aerospace Systems Test Reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

International Nuclear Information System (INIS)

Warinner, D.K.

1983-01-01

This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 10 4 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code

Comparison of the aerospace systems test reactor loss-of-coolant test data with predictions of the 3D-AIRLOCA code

International Nuclear Information System (INIS)

Warinner, D.K.

1984-01-01

This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 10 4 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code. (author)

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

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

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

Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Albrecht, H.; Malinauskas, A.P.

1978-01-01

A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

On the transient pressure build-up in the full pressure safety shell of watercooled nuclear reactors after a loss of coolant accident

International Nuclear Information System (INIS)

Mansfeld, G.

1979-08-01

The thermo-and fluid-dynamic processes in a multichamber full pressure safety containment during a loss of coolant accident have been investigated. Comparison of the calculations carried out with the computer programs, in which ZOCO VI was used as being representative of similar programs, with the experimental results pointed out discrepancies in the determination of time dependent pressure, pressure difference and temperature curves. This led to the development of a new theoretical model and a program COFLOW which pays particular attention to the fluid dynamic processes in the initial phase of a loss of coolant accident. It can also be used to determine the maximum containment pressure towards the end of a loss of coolant accident. Comparison of the COFLOW results with experiments has shown that COFLOW provides a model and a procedure by which the physical processes in a multichamber full pressure safety containment can be simulated satisfactorily

Long-term security of electrical and control engineering equipment in nuclear power stations to withstand a loss of coolant accident

International Nuclear Information System (INIS)

Mueller, H.

1996-01-01

Electrical and control engineering equipment, which has to function even after many years of operation in the event of a fault in a saturated steam atmosphere of 160 C maximum, is essential in nuclear power stations in order to control a loss of coolant accident. The nuclear power station operators have, for this purpose, developed verification strategies for groups of components, by means of which it is ensured that the electrical and control engineering components are capable of dealing with a loss of coolant accident even at the end of their planned operating life. (orig.) [de

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

Transient Analysis for Evaluating the Potential Boiling in the High Elevation Emergency Cooling Units of PWR Following a Hypothetical Loss of Coolant Accident (LOCA) and Subsequent Water Hammer Due to Pump Restart

International Nuclear Information System (INIS)

Husaini, S. Mahmood; Qashu, Riyad K.

2004-01-01

The Generic Letter GL-96-06 issued by the U.S. Nuclear Regulatory Commission (NRC) required the utilities to evaluate the potential for voiding in their Containment Emergency Cooling Units (ECUs) due to a hypothetical Loss Of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) accompanied by the Loss Of Offsite Power (LOOP). When the offsite power is restored, the Component Cooling Water (CCW) pumps restart causing water hammer to occur due to cavity closure. Recently EPRI (Electric Power Research Institute) performed a research study that recommended a methodology to mitigate the water hammer due to cavity closure. The EPRI methodology allows for the cushioning effects of hot steam and released air, which is not considered in the conventional water column separation analysis. The EPRI study was limited in scope to the evaluation of water hammer only and did not provide any guidance for evaluating the occurrence of boiling and the extent of voiding in the ECU piping. This paper presents a complete methodology based on first principles to evaluate the onset of boiling. Also, presented is a methodology for evaluating the extent of voiding and the water hammer resulting from cavity closure by using an existing generalized computer program that is based on the Method of Characteristics. The EPRI methodology is then used to mitigate the predicted water hammer. Thus it overcomes the inherent complications and difficulties involved in performing hand calculations for water hammer. The heat transfer analysis provides an alternative to the use of very cumbersome modeling in using CFD (computational fluid dynamics) based computer programs. (authors)

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

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)

Fission product source from Ignalina NPP in case of loss-of-coolant accidents

International Nuclear Information System (INIS)

Ubonavicius, E.; Rimkevicius, S.

2001-01-01

The release of radioactive materials to the environment is of special importance in the case of any accident at Nuclear Power Plants (NPP). The integrated analysis of thermal-hydraulic parameters behavior and radioactive fission products (FP) transport and deposition in the compartments play an important role in the evaluation of FP release to the environment and determines the irradiation dozes of personnel and public. In this report the transport and the deposition of radioactive material in the Ignalina NPP unit 1 compartments as well as the FP source term to the environment in the case of design basis loss-of-coolant accidents are discussed. The calculation models for the evaluation of FP transport and deposition as well as the results of performed calculations of several accidents at Ignalina NPP are presented. (author)

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.

Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

International Nuclear Information System (INIS)

Soares, Alexandre de Souza

2014-01-01

The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

Simulation of a loss of coolant accident

International Nuclear Information System (INIS)

1987-06-01

An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

Best-estimate analysis of a loss-of-coolant accident in a four-loop US PWR using TRAC-PD2

International Nuclear Information System (INIS)

Ireland, J.R.

1982-01-01

A 200% double-ended cold-leg break loss-of-coolant accident (LOCA) in a typical US pressurized water reactor (PWR) was simulated using the Transient Reactor Analysis Code (TRAC-PD2). The reactor system modeled represented a typical US PWR with four loops and cold-leg emergency-core-cooling systems (ECCS). The calculated peak cladding temperature of 950 K occurred during blowdown and the cladding temperature excursion was terminated at 175 s when complete core quenching occurred. Accumulator flows were initiated at 10 s when the system pressure reached 4.08 MPa, and the refill phase ended at 36 s when the lower plenum refilled. During reflood, both bottom and falling film quench fronts were calculated. Top quenching was caused by entrainment from the lower plenum and lower core regions. The entrained liquid was sufficient to form a small, saturated pool (0.3 m deep) above the upper core support plate. Also, some of the entrained liquid was carried out the hot legs and vaporized in the steam generators. Strong multidimensional effects were calculated in the reactor vessel, particularly with respect to rod quenching

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

Energy Technology Data Exchange (ETDEWEB)

Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

2016-08-15

Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

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

Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability

Science.gov (United States)

Ha, Taesung

A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) 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 accident sequence evaluation procedure (ASEP) approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a time-oriented HRA model (reliability physics model) was applied for the human error probability (HEP) estimation of the core relocation. This model is based on two competing random variables: phenomenological time and performance time. The response surface and direct Monte Carlo simulation with Latin Hypercube sampling were applied for estimating the phenomenological time, whereas the performance time was obtained from interviews with operators. An appropriate probability distribution for the phenomenological time was assigned by statistical goodness-of-fit tests. The human error probability (HEP) for the core relocation was estimated from these two competing quantities: phenomenological time and operators' performance time. The sensitivity of each probability distribution in human reliability estimation was investigated. In order to quantify the uncertainty in the predicted HEPs, a Bayesian approach was selected due to its capability of incorporating uncertainties in model itself and the parameters in that model. The HEP from the current time-oriented model was compared with that from the ASEP approach. Both results were used to evaluate the sensitivity of alternative huinan reliability modeling for the manual core relocation in the LOCA risk model. This exercise demonstrated the applicability of a reliability physics model supplemented with a. Bayesian approach for modeling human reliability and its potential

MABEL-1. A code to analyse cladding deformation in a loss-of-coolant accident

International Nuclear Information System (INIS)

Bowring, R.W.; Cooper, C.A.

1978-06-01

The MABEL-1 code has been written to investigate the deformation, of fuel pin cladding and its effects on fuel pin temperature transients during a loss-of-coolant accident. The code considers a single fuel pin with heated fuel concentric within the cladding. The fuel pin temperature distribution is evaluated using a one-dimensional conduction model with heat transfer to the coolant represented by an input set of heat transfer coefficients. The cladding deformation is calculated using the code CANSWEL, which assumes all strain to be elastic or creep and models the creep under a multi-axial stress system by a spring/dashpot combination undergoing alternate relaxation and elastic strain. (author)

MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

International Nuclear Information System (INIS)

Bowring, R.W.; Cooper, C.A.

1982-04-01

MABEL-2 has been developed to predict the extent of cladding deformation in PWR fuel rods during a loss of coolant accident. The user notes describe how to run MABEL. They include case preparation and input data, the job control language, a description of the output tables available, and aids to debugging. The input data and results for two sample cases are given. (U.K.)

Conservatism of loss-of-coolant accident licensing analysis compared to experimental results and best-estimate calculation

International Nuclear Information System (INIS)

Winkler, F.; Friedmann, P.

1986-01-01

The paper compares results of loss-of-coolant accident licensing analysis with experimental results and results of best-estimate calculations. The large safety margins resulting from the more realistic best-estimate results are used to show the high conservatism inherent in the licensing process of pressurized water reactors. (orig.) [de

Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

International Nuclear Information System (INIS)

Cozzuol, J.M.; Larson, T.K.

1978-01-01

Examination of the U-tubes in the steam generators of some large commercial pressurized water reactors (PWR) has revealed the existence of leakage and in some cases structural weakening of the tubes. This structural weakening enhances the possibility of tubes rupturing during a hypothesized loss-of-coolant accident (LOCA). Considerable interest has been shown in the analysis of tube ruptures concurrent with a hypothesized LOCA since the presence of tube ruptures has the potential to influence the system thermal-hydraulic response and could foreseeably result in a more severe core thermal behavior than might otherwise occur. To experimentally investigate the influence of steam generator tube ruptures on the thermal-hydraulic response of PWR type system, a series of experiments was conducted in the Semiscale Mod-1 system by EG and G Idaho, Inc., for the U.S. Nuclear Regulatory Commission and the Department of Energy. The primary objective of the experiments was to obtain data which could be used to evaluate the influence of the simulated tube ruptures on the system and core thermal-hydraulic response for a range of tube ruptures that was expected to provide the potential for high cladding temperatures in the Semiscale facility. The experiments were conducted assuming a variety in the number of tubes ruptured during large break loss-of-coolant conditions. The number of experiments conducted permitted determination of the range of tube ruptures for which high peak cladding temperatures could result in the Semiscale Mod-1 system. The paper contains a description of the Semiscale Mod-1 system and a discussion of the steam generator tube rupture tests conducted. The experimental results from the test series and the thermal-hydraulic phenomena found to influence the core thermal response during the experiments are discussed

Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

Energy Technology Data Exchange (ETDEWEB)

Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

1995-09-01

The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

Current status of accident analysis for Korean HCCR TBS

Energy Technology Data Exchange (ETDEWEB)

Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Jin, Hyung Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.

Current status of accident analysis for Korean HCCR TBS

International Nuclear Information System (INIS)

Ahn, Mu-Young; Jin, Hyung Gon; Cho, Seungyon; Lee, Dong Won; Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin

2014-01-01

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements

Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

International Nuclear Information System (INIS)

Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B.; Fong, R.W.L.; Coleman, C.E.

1998-08-01

The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHIF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 degrees C. (author)

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.

Reactor Coolant Pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Reactor coolant pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Analysis of an AP600 intermediate-size loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Boyack, B.E.; Lime, J.F. [Los Alamos National Lab., NM (United States)

1995-09-01

A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

International Nuclear Information System (INIS)

Muramatsu, Ken; Akimoto, Masayuki

1982-08-01

THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

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

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

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)

Investigation of a hydrogen mitigation system during large break loss-of-coolant accident for a two-loop pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Dehjourian, Mehdi; Rahgoshay, Mohmmad; Jahanfamia, Gholamreza [Dept. of Nuclear Engineering, Science and Research Branch, Islamic Azad University of Tehran, Tehran (Iran, Islamic Republic of); Sayareh, Reza [Faculty of Electrical and Computer Engineering, Kerman Graduate University of Technology, Kerman (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

2016-10-15

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

International Nuclear Information System (INIS)

Bowring, R.W.; Cooper, C.A.; Haste, T.J.

1982-04-01

MABEL can be used to determine the cladding deformation in a PWR during a LOCA. It takes the results of calculations from other codes to define the initial fuel condition and the transient whole core thermal-hydraulic behaviour. The use of MABEL with input data appropriate to different regions of a reactor core allows an overall picture of coolant channel blockage within the core to be obtained. (U.K.)

Multirods burst tests under loss-of-coolant conditions

International Nuclear Information System (INIS)

Kawasaki, S.; Uetsuka, H.; Furuta, T.

1983-01-01

In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

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)

RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

Energy Technology Data Exchange (ETDEWEB)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

RELAP5/MOD2.5 analysis of the HFBR [High Flux Beam Reactor] for a loss of power and coolant accident

International Nuclear Information System (INIS)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs

On the air coolability of TRIGA reactors following a loss-of-coolant accident

International Nuclear Information System (INIS)

El-Genk, Mohamed S.; Kim, Sung-Ho; Zaki, Galal M.; Foushee, Fabian; Philbin, Jeffrey S.; Schulze, James

1986-01-01

This paper describes the experiments on the air-coolability of a heated rod in a vertical open annulus at near atmospheric pressure. This data can be applied to the coolability of reactor fuel rods that are totally uncovered in a Loss-of-Coolant Accident (LOCA). As a prelude to measuring air coolability of specific core geometries (bundles), heat transfer data was collected for natural convection of atmospheric air in open vertical annuli with an isoflux inner wall and an insulated outer wall (diameter ratios, annulus ratio, of 1.155, 1.33, 1.63, and 12). Although the inner heated tube had the same overall dimensions as the fuel rod in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (3.81 cm o.d. and 55.5 cm long), the heated length was only 36.0 cm rather than the entire 50.5 cm for the ACRR's rods. The test assembly was operated at heat fluxes up to 1.38 W/cm 2 with a corresponding surface temperature of 852 K. The annulus data was extrapolated to an equilibrium surface temperature of 1200 K (as a coolability limit of TRIGA reactors) to provide a qualitative estimate of the coolability of multirod bundles by free convection of atmospheric air. The results suggest that for a typical pitch-to-diameter ratio of 1.12 in the ACRR the decay heat removal level is about 1.0 kW/m. This corresponds to an initial decay power following sustained operations at about 12.5 kW/m in the ACRR. However, because of the uncertainties in duplicating the actual thermal-hydraulic conditions in a multirod bundle using a single rod annulus, the actual coolability of open pool reactors could be different from those suggested in this paper. (author)

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.

TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

International Nuclear Information System (INIS)

Lime, J.F.; Boyack, B.E.

1996-01-01

An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

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

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

International Nuclear Information System (INIS)

Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

2000-01-01

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

International Nuclear Information System (INIS)

Gulden, W.

1977-08-01

The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

Condensing heat transfer following a loss-of-coolant accident

International Nuclear Information System (INIS)

Krotiuk, W.J.; Rubin, M.B.

1978-01-01

A new method for calculating the steam mass condensation energy removal rates on cold surfaces in contact with an air-steam mixture has been developed. This method is based on the principles of mass diffusion of steam from an area of high concentration to the condensing surface, which is an area of low steam concentration. This new method of calculating mass condensation has been programmed into the CONTEMPT-LT Mod 26 computer code, which calculates the pressure and temperature transients inside a light water reactor containment following a loss-of-coolant accident. The condensing heat transfer coefficient predicted by the mass diffusion method is compared to existing semi-empirical correlations and to the experimental results of the Carolinas Virginia Tube Reactor Containment natural decay test. Closer agreement with test results is shown in the calculation of containment pressure, temperature, and heat sink surface temperature using the mass diffusion condensation method than when using any existing semi-empirical correlation

Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1975-01-01

The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

International Nuclear Information System (INIS)

Pointner, W.; Broecker, A.

2012-01-01

The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

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

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

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

Science.gov (United States)

Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

2017-11-01

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

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)

LWR and HTGR coolant dynamics: the containment of severe accidents

International Nuclear Information System (INIS)

Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

1983-07-01

This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

Deformation of PWR cladding following a loss-of-coolant accident

International Nuclear Information System (INIS)

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

1979-07-01

A review is presented of recent experiments to investigate the deformation behaviour of Zircaloy cladding in simulated loss-of-coolant accidents. The behaviour of Zircaloy cladding is shown to be controlled by a complex interaction of metallurgical and heat transfer variables, with the latter having a major influence. There is a significant increase in both diametral strain and the axial extent of deformation in multi-rod compared with single-rod tests. The extent to which this will occur in nuclear-heated tests is not yet known; however, it is expected that the 'smearing' of the gamma-radiation portion of decay heat in such tests will tend to reduce circumferential temperature variations. Opposing this is the influence of the colder control rods in an assembly. The resolution of this dichotomy will require a series of in-reactor multi-rod tests and attendant code development. (author)

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.)

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.

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

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)

Containment parametric analysis for loss of coolant accident

International Nuclear Information System (INIS)

Fabjan, L.

1985-01-01

Full text: This paper presents parametric analysis of double containment response to LOCA using CONTEMPT-LT/28 code. The influence of the active and passive heat sinks on thermodynamic parameters in the containment after big and small LOCA was considered. (author)

Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

International Nuclear Information System (INIS)

Ruger, C.J.; Luckas, W.J. Jr.

1989-03-01

Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

RETRAN analysis of inter-system LOCA within the primary coolant pump

International Nuclear Information System (INIS)

Gangadharan, A.; Pratt, G.F.

1992-01-01

One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

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)

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Energy Technology Data Exchange (ETDEWEB)

Hadad, Kamal; Esmaeili-Sanjavanmareh, Mansour [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-05-15

PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

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)

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

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)

Analysis of fuel behaviour after loss-of-coolant accident with the TESPA-code

International Nuclear Information System (INIS)

Keusenhoff, J.

1981-01-01

After a loss-of-coolant accident fuel rods go through a phase of high temperature and differential pressure before quenching and initiation of long term cooling. For licensing purpose the highest cladding temperature and the coolability of the core is of interest. The highest temperature is evaluated by a hot channel calculation with conservative assumptions. It gives little information about the status of the entire core. Therefore more detailed information is necessary. TESPA is a fast running code, which uses best-estimate assumptions, considers statistical uncertainties in the input parameters and calculates clad ballooning and rupture. The code is a usefull tool for calculation of channel blockage and cladding rupture

Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

OpenAIRE

Bo Cheng; Young-Jin Kim; Peter Chou

2016-01-01

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

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%.

Investigation of loss of coolant accidents in pressurized water reactors using the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method for considering of uncertainties in TRACE

International Nuclear Information System (INIS)

Sporn, Michael; Hurtado, Antonio

2016-01-01

Loss of coolant accident must take uncertainties with potentially strong effects on the accident sequence prediction into account. For example, uncertainties in computational model input parameters resulting from varying geometry and material data due to manufacturing tolerances or unavailable measurements should be considered. The uncertainties of physical models used by the software program are also significant. In this paper, use of the ''Dynamic Best-Estimate Safety Analysis'' (DYBESA) method to quantify the uncertainties in the TRACE thermal-hydraulic program is demonstrated. For demonstration purposes loss of coolant accidents with breaks of various types and sizes in a DN 700 reactor coolant pipe are used as an example Application.

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)

Corrosion and solubility in a TSP-buffered chemical environment following a loss of coolant accident: Part 2 – Zinc

International Nuclear Information System (INIS)

Pease, David; LaBrier, Daniel; Ali, Amir; Blandford, Edward D.; Howe, Kerry J.

2016-01-01

Highlights: • Zinc release is limited to less than 1 mg/L in TSP-buffered solution under a variety of conditions (pH, temperature, zinc source). • Zinc release in high-temperature non-TSP-buffered environment is approximately 25 mg/L. • Long-term zinc release is controlled by passivation (without TSP) and zinc solubility (with TSP). • Precipitation and solubility of zinc phosphate limit the release of zinc. - Abstract: Bench experiments were conducted to investigate the effect of the presence of trisodium phosphate (TSP) on the corrosion and release of zinc from metallic zinc-bearing surfaces under conditions representative of the containment pool following a postulated loss of coolant accident (LOCA) at a nuclear power generating facility. The experiments showed that in non-buffered (acidic) environments, measurable quantities of zinc are released from zinc-bearing surfaces. Precipitation and solubility of phosphate-based corrosion products, such as zinc phosphate, limit the release of zinc from zinc-bearing surfaces. These experiments have found that under a variety of conditions, including variations of temperature, pH, and across different zinc-bearing surfaces, the release of zinc into solution is limited to <1 mg/L when phosphate is present. When phosphate is not present, zinc release is instead bounded by a markedly higher saturation limit which is a strong function of the solution temperature.

Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

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)

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)

Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

International Nuclear Information System (INIS)

Wang, T.-C.; Wang, S.-J.; Teng, J.-T.

2005-01-01

This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

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)

Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

International Nuclear Information System (INIS)

Zhang Yanzhao; Zhang Fan; Zhao Xinwen; Zheng Yingfeng

2013-01-01

Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

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)

Analytical and experimental assessment of TVS-2006 fuel assembly thermal-mechanical shape deformation at temperature modeling of a loss-of-coolant accident

International Nuclear Information System (INIS)

Afanasiev, A.; Semishkin, V.; Makarov, V.; Matvienko, I.; Puzanov, D.

2015-01-01

Full or partial core drying-out takes place in loss-of-coolant accidents, which leads to worsening of heat removal from the fuel rods. Depending on the accident scenario the fuel rod cladding temperature can be in a wide range from 350 to 1200°C. It is worth mentioning, that the length of the process can considerably affect the fuel rod cladding loadcarrying capacity and the FA structure as a whole, and in the long run it defines the radiation consequences of the accident and the possibility of postaccident core disassembly at low cost. Most experiments staged of late were devoted to a study of FA behaviour in the temperature range 800-900°C of α→β phase transition that is characterized by a sharp increase in the rate of zirconium alloy creep which leads to fuel rod cladding ballooning and loss of their tightness within a short period of time. The 600-700°C temperature range turned out to be less investigated whereas this is the range where the change of zirconium alloy mechanical properties is also observed but only with the retention of α-phase. The tests of a full-scale FA dummy with the skeleton of guide tubes and spacer grids connected by friction forces, carried out at the testing facility of JSC OKB “GIDROPRESS”, were devoted to a study of FA behaviour in this temperature range. The model was heated up with hot air to 650°C for 6 hours. The tests ended with fuel rod cladding ballooning due to gauge pressure and shape deformation. No loss of fuel rod cladding integrity was observed. Therefore, a conclusion can be made that a long-time core holdup at the parameters implemented at the test facility is permitted and the deformations of the FA structure do not lead to the damage that could considerably complicate the core disassembly. The test results were used for the verification of the calculational model of FA TVS-2006 structure with a welded skeleton by ANSYS code. On the basis of the verified calculational model a calculational model was

Status Report on Spent Fuel Pools under Loss-of-Cooling and Loss-of-Coolant Accident Conditions - Final Report

International Nuclear Information System (INIS)

Adorni, M.; Esmaili, H.; Grant, W.; Hollands, T.; Hozer, Z.; Jaeckel, B.; Munoz, M.; Nakajima, T.; Rocchi, F.; Strucic, M.; ); Tregoures, N.; Vokac, P.; Ahn, K.I.; Bourgue, L.; Dickson, R.; Douxchamps, P.A.; Herranz, L.E.; Jernkvist, L.O.; Amri, A.; Kissane, M.P.; )

2015-01-01

scenarios, past accidents and precursor events; Chapter 4: Behaviour of spent fuel facilities during the Fukushima Daiichi accident; Chapter 5: Accident phenomenology; Chapter 6: Experiments with relevance to SFP cooling accidents; Chapter 7: Simulation tools; Chapter 8: Conclusions and recommendations; The present report summarizes results of experiments and computational analyses carried out to date to gain understanding of phenomena with significance to SFP cooling accidents. Considering that some knowledge gaps currently exist and that ongoing and planned research projects are expected to produce results that will hopefully narrow these gaps within the foreseeable future, it is recommended that: - a CSNI state-of-the-art report on SFP loss-of-cooling and loss-of-coolant accidents is written as the results of these research projects become available; - a follow-on activity is launched on SFP combining probability of SFP accidents, which was beyond the scope of this document, and mitigation strategies