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Sample records for small break loca

  1. 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)

  2. 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)

  3. 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

  4. 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)

  5. Operator reliability analysis during NPP small break LOCA

    International Nuclear Information System (INIS)

    Zhang Jiong; Chen Shenglin

    1990-01-01

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

  6. 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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    1991-01-01

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

  11. 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

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

    International Nuclear Information System (INIS)

    Suh, Jong Tae; Bae, Kyoo Hwan

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Lee, Jong-Ho; Chang, Soon-Heung

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  15. Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

    International Nuclear Information System (INIS)

    Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

    2004-01-01

    The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

  16. 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%.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  19. 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

  20. 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

  1. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A.; Elkin, I.V.; Pylev, S.S.

    2005-01-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  2. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A. [Elektrogorsk Research and Engineering Center, EREC, Bezymiannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Elkin, I.V.; Pylev, S.S. [NSI RRC ' Kurchatov Institute' , Kurchatov Sq., 1, Moscow, 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  3. 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

  4. 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

  5. Comparison of the DVI line break LOCA with the equivalent cold leg break with the ATLAS facility

    International Nuclear Information System (INIS)

    Choi, K. Y.; Cho, S.; Kang, K. H.; Park, H. S.; Kim, Y. S.; Baek, W. P.

    2010-01-01

    The APR1400 (Advanced Power Reactor, 1400 MWe) adopts a DVI (Direct Vessel Injection) method for ECC (Emergency Core Cooling) water delivery rather than a conventional CLI (Cold Leg Injection) method as an advanced safety feature. The break scenario of one DVI nozzle is taken into account in the small break LOCA analysis. Transient behavior during the DVI line breaks needs to be investigated and compared with the equivalent break on the cold leg. An 8.5-inch double-ended break of one DVI nozzle was simulated with the ATLAS, and a counterpart test for the DVI break was performed at the cold leg with the equivalent break size for comparison. This comparison will contribute to enhancing a comprehensive understanding of the thermal hydraulic behavior during transients. A constructed integral effect database is also used to validate the existing conservative safety analysis methodology and to develop a best-estimate safety analysis methodology for small-break LOCAs. A post-test calculation was performed with a best-estimate safety analysis code, MARS 3.1, in order to examine its prediction capability and to identify any code deficiencies for thermal hydraulic phenomena occurring during the transient. (authors)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  7. Estimation of 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.

  8. Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

    International Nuclear Information System (INIS)

    Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

    1988-12-01

    A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

  9. 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.

  10. Temporary core liquid level depression during cold-leg small-break LOCA effect of break size and power level

    International Nuclear Information System (INIS)

    Koizumi, Y.; Kumamaru, H.; Mimura, Y.; Kukita, Y.; Tasaka, K.

    1989-01-01

    Cold-leg small break LOCA experiments (0.5-10% break) were conducted at the large scale test facility (LSTF), a volumetrically-scaled (1/48) simulator of a PWR, of the ROSA-IV Program. When a break area was less than 2.5% of the scaled cold-leg flow area, the core liquid level was temporarily further depressed to the bottom elevation of the crossover leg during the loop seal clearing early in the transient only by the manometric pressure balance since no coolant remained in the upper portion of the primary system. When the break size was larger than 5%, the core liquid level was temporarily further depressed lower than the bottom elevation of the crossover leg during the loop seal clearing since coolant remained at the upper portion of the primary system; the steam generator (SG) U-tube upflow side and the SG inlet plenum, due to counter current flow limiting by updrafting steam while the coolant drained. The amount of coolant trapped there was dependent on the vapor velocity (core power); the larger the core power, the lower the minimum core liquid level. The RELAP5/MOD2 code reasonable predicted phenomena observed in the experiments. (orig./DG)

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

    International Nuclear Information System (INIS)

    1996-07-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. 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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  16. Small break loss of coolant accidents: Bottom and side break

    International Nuclear Information System (INIS)

    Hardy, P.G.; Richter, H.J.

    1987-01-01

    A LOCA can be caused, e.g. by a small break in the primary cooling system. The rate of fluid escaping through such a break will define the time until the core will be uncovered. Therefore the prediction of fluid loss and pressure transient is of major importance to plan for timely action in response to such an event. Stratification of the two phases might be present upstream of the break, thus, the location of the break relative to the vapor-liquid interface and the overall upstream fluid conditions are relevant for the calculation of fluid loss. Experimental results and analyses are presented here for small breaks at the bottom or at the side of a small pressure vessel. It was found that in such a case the onset of the so-called ''vapor pull through'' is important but swelling at sufficient depressurization rates of the liquid due to flashing is also of significance. It was also discovered that in the bottom break the flow rate is strongly dependent on the break entrance quality of the vapour-liquid mixture. The side break can be treated similarly to the bottom break if the interface level is above the break. The analyses developed on the basis of experimental observations showed reasonable agreement of predicted and measured pressure transients. It was possible to calculate the changing interface level and mixture void fraction history in a way compatible with the behavior observed during the experiments. Even though the experiments were performed at low pressures, this work should help to get a better understanding of physical phenomena occurring in a full scale small break LOCA. (orig./HP)

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

    International Nuclear Information System (INIS)

    1992-01-01

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

  18. 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

  19. Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

    International Nuclear Information System (INIS)

    Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

    1997-01-01

    The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  1. 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

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

    1988-07-01

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

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

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  6. 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)

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

    International Nuclear Information System (INIS)

    1992-02-01

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

  8. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

    Directory of Open Access Journals (Sweden)

    Li Yuquan

    2017-02-01

    Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

  9. Comparative experiments to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Li, YuQuan; Hao, Botao; Zhong, Jia; Wan Nam [State Nuclear Power Technology R and D Center, South Park, Beijing Future Science and Technology City, Beijing (China)

    2017-02-15

    The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME)—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative

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

    International Nuclear Information System (INIS)

    Chen Yuzhen

    1992-01-01

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

  11. Appendix S-NH-1 and S-NH-2 of the experiment operating specification for the semiscale MOD-2C small break LOCA without HPI experiment series

    International Nuclear Information System (INIS)

    Owca, W.A.

    1985-10-01

    This document is Appendix S-NH--1 and S-NH-2 of the Experiment Operating Specification (EOS) for the Small Break LOCA without high pressure injection (HPI) series. It contains detailed information on the S-NH-1 and S-NH-2 experiment operation and facility configuration necessary to meet the series objectives stated in the main EOS body. 14 refs., 17 figs

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

    International Nuclear Information System (INIS)

    Kamata, Shinya; Kubo, Kazuo

    2004-01-01

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

  13. Small break LOCA analysis for YGN 5 and 6 RCP trip strategy in power mode operation

    International Nuclear Information System (INIS)

    Kim, Tech Mo; Choi, Han Rim

    2001-01-01

    A continued operation of Reactor Coolant Pumps(RCPs) during a Small Break Loss of Coolant Accident(SBLOCA) in all operation mode may increase unnecessary inventory loss from the Reactor Coolant System(RCS) causing a severe core uncovery which might lead to fuel failure. After Three Mile Island Unit 2(TMI-2) accident, the Combustion Engineering Owner Group(CEOG) developed RCP trip strategy called 'Trip-Two/Leave-Two' (T2/L2). The T2/L2 RCP trip strategy consists of tripping the first two RCPs on low RCS pressure and then tripping the remaining two RCPs if a LOCA has occurred. This analysis demonstrates the inherent safety of RCP trip strategy during an SBLOCA for Youggwang Nuclear Power Plant Unit 5 and 6(YGN 5 and 6). The trip setpoint of the first two RCPs for YGN 5 and 6 is calculated to be 1721 psia in pressurizer pressure based on the limiting SBLOCA with 0.15 ft 2 break size in the hot leg. The analysis results show that YGN 5 and 6 can maintain the core coolability even if the operator fails to trip the second two RCPs or trips at the worst time of minimum liquid inventory

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. 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)

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

    International Nuclear Information System (INIS)

    2012-08-01

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

  17. 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

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

    International Nuclear Information System (INIS)

    1992-01-01

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

  19. Analysis of small break loss of coolant accident for Chinese CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

    2016-05-15

    This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

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

  2. LOCA analysis of the IRIS reactor

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. Scenarios simulation of severe accident type small loss of coolant (Loca), with the code MELCOR version 2.1 for the nuclear power plant of Laguna Verde

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

    2013-10-01

    In this work was carried out the analysis of two scenarios of the accident type with loss of coolant in a recirculation loop for a break with smaller ares to 0.1 ft 2 (4.6 cm 2 ), which is classified according to their size like small Loca. The first simulated scenario was a small Loca without action of the emergency coolant injection systems, and the second was a small Loca with only the available system LPCS. This design base accident was taken into account for its relevance with regard to the damage to the core and the hydrogen generation. Was also observed and analyzed the response of the action of the ECCS that depend of the loss of coolant reason and this in turn depends of the size and type of the pipe break. The specified scenarios were simulated by means of the use of MELCOR model for the nuclear power plant of Laguna Verde that has the Comision Nacional de Seguridad Nuclear y Salvaguardias. (Author)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    2003-01-01

    The objective of the Workshop was to facilitate an exchange of information on a topic, which could potentially impact both the operation of current reactors and the design of future reactors. A number of OECD countries were actively working in this area at the moment. Regulators, Researchers and Industry representatives needed to exchange information on the current regulation and technical issues associated with the Large Break LOCA (LB-LOCA), and to further discuss rationales and motives which could lead to a redefinition of the LB- LOCA. The focus was on design and safety implications. Policy issues were not discussed but the workshop provided technical inputs for policy makers. The workshop covered different reactor designs (CANDUs, VVERs, LWRs). The workshop was articulated over three questions: 1. What drives the need to redefine the LB-LOCA? There was a consensus on well founded drivers for the redefinition and on observations that, if the change is made right, it can enhance the safety of nuclear power plants and also reduce the costs of power production. Three papers were presented. In the first paper, Mr Bajorek (USNRC) discussed the redefinition of the LB-LOCA in the context of risk informing their regulations. Mr. Bajorek emphasised that redefining LB-LOCA is being considered by US NRC from a risk perspective to improve the safety focus and that regulators should better focus on safety and risk contributors and thereby formulate the regulations to better use available resources. He added that the present LOCA definition has not only a great impact on the plant design but also on operating limits according to the Technical Specifications as well as on testing conditions. In the second paper, Mr Pietrangelo (NEI, USA) presented a paper on the need to redefine the large break LOCA from the industrial viewpoints. He emphasised that the strong leadership commitments by both NRC and industry are necessary, and that redefining LB- LOCA is central to risk

  6. 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)

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

    Directory of Open Access Journals (Sweden)

    Cesare Frepoli

    2008-01-01

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

  8. 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

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

    International Nuclear Information System (INIS)

    Brittain, I.; Aksan, S.N.

    1990-08-01

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

  10. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2010-02-01

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  15. 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)

  16. 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)

  17. RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

    2016-08-15

    Highlights: • RELAP5/SCDAPSIM model of AP1000 has been developed. • Analysis involves a LBLOCA (double ended guillotine break) study in cold leg. • Results are compared with those of WCOBRA–TRAC and TRACE. • Concluded that PCT does not violate the safety criteria of 1477 K. - Abstract: The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved

  18. 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

  19. 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)

  20. HCCR TBS LOCA and ICE into small confined volume

    International Nuclear Information System (INIS)

    Jin, Hyung Gon; Ahn, Mu-Young

    2016-01-01

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

  1. Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes

    International Nuclear Information System (INIS)

    Papini, Davide; Grgic, Davor; Cammi, Antonio; Ricotti, Marco E.

    2011-01-01

    Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones. In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes). The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. BWR recirculation loop discharge line break LOCA tests with break areas of 50 and 100% assuming HPCS failure at ROSA-III test facility

    International Nuclear Information System (INIS)

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

    1985-03-01

    This report presents the experimental results of RUN 962 and RUN 963 in ROSA-III program, which are 50 and 100 % break LOCA tests at the BWR recirculation pump discharge line, respectively. The ROSA-III test facility simulates a volumetrically scaled (1/424) BWR system and has four half-length electrically heated fuel bundles, two active recirculation loops, three types of ECCSs and steam and feedwater systems. The experimental data of RUN 962 and RUN 963 were compared with those of RUN 961, a 200 % discharge line break test to study the break area effects on the transient thermal hydraulic phenomena. The least flow areas at the jet pump drive nozzles and recirculation pump discharge nozzle in the broken recirculation loop limitted the discharge flows from the pressure vessel and the depressurization rate in the 100 and 200 % break tests, whereas the least flow area at break nozzle limitted the depressurization rate in the 50 % break test. The highest PCT was observed in the 50 % break test among the three tests. (author)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Post-test analysis of semiscale large-break test S-06-3 using TRAC-PF1

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1982-01-01

    The Transient Reactor Analysis Code (TRAC) is an advanced systems code for light-water-reactor accident analysis. The code was developed originally to analyze large-break loss-of-coolant accidents (LOCAs) and running time was not a primary development criterion. TRAC-PF1 was developed because increased application of the code to long transients such as small-break LOCAs required a faster-running code version. Although developed for long transients, its performance on large-break transients is still important. This paper assesses the ability of TRAC-PF1 to predict large-break-LOCA Test S-06-3 conducted in the Semiscale Mod-1 facility

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

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

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

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

    International Nuclear Information System (INIS)

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

    1986-03-01

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

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

    International Nuclear Information System (INIS)

    Frepoli, Cesare; Oriani, Luca

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    1989-01-01

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

  10. 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

  11. 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)

  12. Analysis, by Relap5 code, of boron dilution phenomena in a Small Break Loca Transient, performed in PKL III E 2.2 test

    International Nuclear Information System (INIS)

    Rizzo, G.; Vella, G.

    2007-01-01

    The present work is finalized to investigate the E2.2 thermal-hydraulics transient of the PKL III facility, which is a scaled reproduction of a typical German PWR, operated by FRAMATOME-ANP in Erlangen, Germany, within the framework of an international cooperation (OECD/SETH project). The main purpose of the project is to study boron dilution events in Pressurized Water Reactors and to contribute to the assessment of thermal-hydraulic system codes like Relap5. The experimental test PKL III E2.2 investigates the behavior of a typical PWR after a Small Break Loss Of Coolant Accident (SB-LOCA) in a cold leg and an immediate injection of borated water in two cold legs. The main purpose of this work is to simulate the PKL III test facility and particularly its experimental transient by Relap5 system code. The adopted nodalization, already available at Department of Nuclear Engineering (DIN), has been reviewed and applied with an accurate analysis of the experimental test parameters. The main result relies in a good agreement of calculated data with experimental measures for a number of main important variables. (author)

  13. New theoretical model for two-phase flow discharged from stratified two-phase region through small break

    International Nuclear Information System (INIS)

    Yonomoto, Taisuke; Tasaka, Kanji

    1988-01-01

    A theoretical and experimental study was conducted to understand two-phase flow discharged from a stratified two-phase region through a small break. This problem is important for an analysis of a small break loss-of-coolant accident (LOCA) in a light water reactor (LWR). The present theoretical results show that a break quality is a function of h/h b , where h is the elevation difference between a bulk water level in the upstream region and break and b the suffix for entrainment initiation. This result is consistent with existing eperimental results in literature. An air-water experiment was also conducted changing a break orientation as an experimental parameter to develop and assess the model. Comparisons between the model and the experimental results show that the present model can satisfactorily predict the flow rate and the quality at the break without using any adjusting constant when liquid entrainment occurs in a stratified two-phase region. When gas entrainment occurs, the experimental data are correlated well by using a single empirical constant. (author)

  14. UPTF/TEST10B/RUN081, Steam/Water Flow Phenomena Reflood PWR Cold Leg Break LOCA

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of test facility: The Upper Plenum Test Facility (UPTF) is a geometrical full-scale simulation of the primary system of the four-loop 1300 MWe Siemens/KWU pressurized water reactor (PWR) at Grafenrheinfeld. The test vessel, upper plenum and its internals, downcomer, primary loops, pressurizer and surge line are replicas of the reference plant. The core, coolant pumps, steam generators and containment of a PWR are replaced by simulators which simulate the boundary and initial conditions during end-of-blowdown, refill and reflood phase following a loss-of-coolant accident (LOCA) with a hot or cold leg break. The break size and location can be simulated in the broken loop. The emergency core coolant (ECC) injection systems at the UPTF are designed to simulate the various ECC injection modes, such as hot leg, upper plenum, cold leg, downcomer or combined hot and cold leg injection of different ECC systems of German and US/Japan PWRs. Moreover, eight vent valves are mounted in the core barrel above the hot leg nozzle elevation for simulation of ABB and B and W PWRs. The UPTF primary system is divided into the investigation and simulation areas. The investigation areas, which are the exact replicas of a GPWR, consist of the upper plenum with internals, hot legs, cold legs and downcomer. The realistic thermal-hydraulic behavior in the investigation areas is assured by appropriate initial and boundary conditions of the area interface. The boundary conditions are realized by above mentioned simulators, the setup and the operation of which are based on small-scale data and mathematical models. The simulation areas include core simulator, steam generator simulators, pump simulators and containment simulator. The steam production and entrainment in a real core during a LOCA are simulated by steam and water injection through the core simulator. 2 - Description of test: Investigation of steam/water flow phenomena at the upper tie plate and in the upper plenum and

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

    International Nuclear Information System (INIS)

    Araya, F.; Akimoto, M.

    1985-01-01

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

  16. 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.)

  17. Blind-blind prediction by RELAP5/MOD1 for a 0.1% very small cold-leg break experiment at ROSA-IV large-scale test facility

    International Nuclear Information System (INIS)

    Koizumi, Y.; Kumamaru, H.; Kukita, Y.; Kawaji, M.; Osakabe, M.; Schultz, R.R.; Tanaka, M.; Tasaka, K.

    1986-01-01

    The large-scale test facility (LSTF) of the Rig of Safety Assessment No. 4 (ROSA-IV) program is a volumetrically scaled (1/48) pressurized water reactor (PWR) system with an electrically heated core used for integral simulation of small break loss-of-coolant accidents (LOCAs) and operational transients. The 0.1% very small cold-leg break experiment was conducted as the first integral experiment at the LSTF. The test provided a good opportunity to truly assess the state-of-the-art predictability of the safety analysis code RELAP5/MODI CY18 through a blind-blind prediction of the experiment since there was no prior experience in analyzing the experimental data with the code; furthermore, detailed operational characteristics of LSTF were not yet known. The LOCA transient was mitigated by high-pressure charging pump injection to the primary system and bleed and feed operation of the secondary system. The simulated reactor system was safely placed in hot standby condition by engineered safety features similar to those on a PWR. Natural circulation flow was established to effectively remove the decay heat generated in the core. No cladding surface temperature excursion was observed. The RELAP5 code showed good capability to predict thermal-hydraulic phenomena during the very small break LOCA transient. Although all the information needed for the analysis by the RELAP5 code was obtained solely from the engineering drawings for fabrication and the operational specifications, the code predicted key phenomena satisfactorily

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  19. Scientific design of the test facility for the KNGR DVI line small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Park, Choon Kyung; Jun, Hyung Gil; Cho, Seok; Kwon, Tae Soon; Song, Chul Hwa; Kim, Jung Taek

    1999-03-01

    Scientific design of the experimental facility (OASIS) for the KNGR (Korea Next Generation Reactor) DVI line SB-LOCA simulation is carried out. Main purpose of the OASIS is to produce thermal-hydraulic data base for determining the best location of the DVI (Direct Vessel Injection) injection nozzle of the KNGR as well as verifying its design performance in view of the ECCS (Emergency Core Cooling System) effectiveness. The experimental facility is designed based on the Ishii's three-level scaling law. The facility has 1/4 height and 1/341 area scaling ratio. It corresponds to the volume scale of 1/1364. The power scaling is 1/682 and the system pressure is prototypic. The OASIS consists of a core, a downcomer, two steam generators, two pump simulators, a break simulator, a collection tank, primary piping as well as a circulation pump for initial test condition. Each component is designed based on the Ishill's global scaling and boundary flow scaling of mass, energy and momentum. In addition, local phenomena scaling is carried out for the design of major components to preserve key local phenomena in each component. Most of the key phenomena are well preserved in the OASIS. However, the local scaling analysis shows that distortions of the void fraction and mixture level can not be avoided in the core. It comes from the basic features of the Ishill's scaling law in case of the reduced-height simulation. However, it is expected that these distortions will be analyzed properly by a best estimate system analysis code. (Author). 22 refs., 20 tabs., 25 figs.

  20. Angra 2 small break LOCA flow regime identification through RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Rocha, Marcelo da Silva; Sabundjian, Gaiane; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Torres, Walmir Maximo; Conti, Thadeu das Neves; Macedo, Luiz Alberto; Umbehaun, Pedro Ernesto; Mesquita, Roberto Navarro de; Masotti, Paulo Henrique Ferraz, E-mail: msrocha@ipen.br, E-mail: gdjian@ipen.br, E-mail: abelchior@ipen.br, E-mail: delvonei@ipen.br, E-mail: wmtorres@ipen.br, E-mail: tnconti@ipen.br, E-mail: lamacedo@ipen.br, E-mail: umbehaun@ipen.br, E-mail: s, E-mail: rnavarro@ipen.br, E-mail: pmasotti@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2012-07-01

    The purpose of this paper is to identify the flow regimes in the core of Angra 2 nuclear reactor with RELAP5/MOD3.2.gamma code (RELAP5, 2001). The postulated accident is the loss of coolant through a small break in the primary circuit (SBLOCA), which is described in Chapter 15 of the Final Safety Analysis Report of Angra 2 - FSAR (ETN, 2006). As the primary circuit pressure decreases due to the loss of coolant, several alternating two phase flow regimes are established in the primary circuit. This paper analyses the coolant two-phase flow behavior in the nuclear reactor core during the postulated accident. (author)

  1. 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.

  2. 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)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  5. Prediction of Counter-Current Flow Limitation at Hot Leg Pipe During a Small-Break Loca

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, H.Y. [Korea Electric Power Research Institute, Taejeon (Korea)

    2001-07-01

    The possibility of hot leg flooding during reflux condensation cooling after a small-break loss-of-coolant accident in a nuclear power plant is evaluated. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The effect of initial water level to counter-current flow limitation is taken into account. It is predicted that the hot leg flooding is precluded when all steam generators are available for heat removal. It is also shown the both hot leg flooding and SG flooding are possible under the operation of one steam generators. Therefore, it can be said that the occurrence of hot leg flooding under reflux condensation cooling is possible when the number of steam generators available for heat removal is limited. (author). 15 refs., 15 figs., 3 tabs.

  6. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

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

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  10. Counter-current flow limitation at hot leg pipe during reflux condensation cooling after small-break LOCA

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Ha, Sang Jun; Jo, Yung Jo; Jun, Hwang Yong

    1999-01-01

    The possibility of hot leg flooding is evaluated in case of a small-break loss-of-coolant accident in Korean Next Generation Reactor (KNGR) operating at the core power of 3983 MW normally. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13 %, 1.02 % and 10.19 % cold leg break. The calculated results are compared with the existing flooding correlations. It is predicted that the hot leg flooding is excluded when two steam generators are available. It is also shown that the possibility of hot leg flooding under the operation with one steam generator is very low. Therefore, it can be said that the occurrence of hot leg flooding is unexpected when the reflux condensation cooling is maintained in steam generator tubes

  11. 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

  12. Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements

    International Nuclear Information System (INIS)

    Wilkowski, G.; Rudland, D.; Wolterman, R.; Krishnaswamy, P.; Scott, P.; Rahman, S.; Fairbanks, C.

    2002-01-01

    The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, A cceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors . These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from the

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. 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

  15. Analysis of ATLAS 6-inch cold leg break simulation with MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Se Yun; Jun, Hwang Yong; Ha, Sang Jun [Korea Electric Power Company, Daejeon (Korea, Republic of)

    2011-05-15

    A Domestic Standard Problem (DSP) exercise using ATLAS facility has been organized by KAERI. As the second DSP exercise, the 6-inch cold leg bottom break was determined. This experiment is the counterpart test to the DVI line break to verify the safety performance of the DVI method over the traditional CLI method. Compared with the large break LOCA, the phases of the small break LOCA prior to core recovery occur over a long period. The blowdown, natural circulation, loop seal clearance, boil-off, and core recovery phase should be investigated minutely with relevant models of safety analysis codes in order to predict these thermal hydraulic phenomena correctly. To investigate the ECC bypass phenomena, a finer study on the thermalhydraulic behavior in upper annulus downcomer was carried out

  16. 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)

  17. Performance Evaluation of SMART Passive Safety System for Small Break LOCA Using MARS Code

    International Nuclear Information System (INIS)

    Chun, Ji Han; Lee, Guy Hyung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

    2013-01-01

    SMART has significantly enhanced safety by reducing its core damage frequency to 1/10 that of a conventional nuclear power plant. KAERI is developing a passive safety injection system to replace the active safety injection pump in SMART. It consists of four trains, each of which includes gravity-driven core makeup tank (CMT) and safety injection tank (SIT). This system is required to meet the passive safety performance requirements, i.e., the capability to maintain a safe shutdown condition for a minimum of 72 hours without an AC power supply or operator action in the case of design basis accidents (DBAs). The CMT isolation valve is opened by the low pressurizer pressure signal, and the SIT isolation valve is opened at 2 MPa. Additionally, two stages of automatic depressurization systems are used for rapid depressurization. Preliminary safety analysis of SMART passive safety system in the event of a small-break loss-of-coolant accident (SBLOCA) was performed using MARS code. In this study, the safety analysis results of a guillotine break of safety injection line which was identified as the limiting SBLOCA in SMART are given. The preliminary safety analysis of a SBLOCA for the SMART passive safety system was performed using the MARS code. The analysis results of the most limiting SI line guillotine break showed that the collapsed liquid level inside the core support barrel was maintained sufficiently high above the top of core throughout the transient. This means that the passive safety injection flow from the CMT and SIT causes no core uncovery during the 72 hours following the break with no AC power supply or operator action, which in turn results in a consistent decrease in the fuel cladding temperature. Therefore, the SMART passive safety system can meet the passive safety performance requirement of maintaining the plant at a safe shutdown condition for a minimum of 72 hours without AC power or operator action for a representing accident of SBLOCA

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

    International Nuclear Information System (INIS)

    1988-12-01

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

  19. Replacement divider plate performance under LOCA loading

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  20. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  2. Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    Rychkov, M.; Chikkanagoudar, U.; Sehgal, B.R.

    2004-01-01

    A RELAP5 model for the analysis of the PSB-VVER test facility was developed by EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, we have modified the PSB-VVER facility's RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5's calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the '11% UP LOCA' test data, the RELAP5/MOD3.2 model was used for a so-called 'blind' transient calculation of the test '2*25% HL LOCA' and the results obtained were compared with the experimental data provided after the calculation. From the results of the qualitative and quantitative comparison of the 2 test calculations and the experimental data, we can state that the RELAP5/MOD3.2 code gives a satisfactory modeling of the PSB-VVER facility' thermal hydraulic phenomena

  3. 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)

  4. Simulation with the MELCOR code of two severe accident sequences, Station Blackout and Small Break LOCA, for the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Valle Cepero, Reinaldo

    2004-01-01

    The results of the PSA-I applied to the Atucha I nuclear power plant (CNA I) determine the accidental sequences with the most influence related to the probability of the core reactor damage. Among those sequences are include, the Station Blackout and lost of primary coolant, combine with the failure of the emergency injection systems by pipe breaks of diameters between DN100 - DN25 or equivalent areas, Small LOCA. This paper has the objective to model and analyze the behavior of the primary circuit and the pressure vessel during the evolution of those two accidental sequences. It presented a detailed analysis of the main phenomena that occur from the initial moment of the accident to the failure moment of the pressure vessel and the melt material fall to the reactor cavity. Two sequences were taken into account, considering the main phenomena (core uncover, heating, fuel element oxidation, hydrogen generation, degradation and relocation of the melt material, failure of the support structures, etc.) and the time of occurrence, of those events will be different, if it is considered that both sequences will be developed in different scenarios. One case is an accident with the primary circuit to a high pressure (Station Blackout scenario) and the other with a early primary circuit depressurization due to the lost of primary coolant. For this work the MELCOR 1.8.5 code was used and it allows within a unified framework to modeling an extensive spectrum of phenomenology associated with the severe accidents. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-06-15

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

  6. Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

    International Nuclear Information System (INIS)

    Lim, Sang Gyu; Lee, Suk Ho; Yu, Keuk Jong; Kim, Han Gon; Lee, Jae Yong

    2013-01-01

    The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft 2 ∼0.02ft 2 (DVI) 0.5ft 2 ∼0.02ft 2 (hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA

  7. Cold-Leg Small Break LOCA Analysis of APR1400 Plant Using a SPACE/sEM Code

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang Gyu; Lee, Suk Ho; Yu, Keuk Jong; Kim, Han Gon; Lee, Jae Yong [Central Research Institute, KHNP, Ltd., Daejeon (Korea, Republic of)

    2013-10-15

    The Small Break Loss-of-Coolant Accident (SBLOCA) evaluation methodology (EM) for APR1400, called sEM, is now being developed using SPACE code. SPACE/sEM is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. Major required and acceptable features of the evaluation models are described as below. - Fission product decay : 1.2 times of ANS97 decay curve - Critical flow model : Henry-Fauske Moody two phase critical flow model - Metal-Water reaction model : Baker-Just equation - Critical Heat Flux (CHF) : B and W, Barnett and Modified Barnett correlation - Post-CHF : Groeneveld 5.7 film boiling correlation A series of test matrix is established to validate SPACE/sEM code in terms of major SBLOCA phenomena, e.g. core level swelling and boiling, core heat transfer, critical flow, loop seal clearance and their integrated effects. The separated effect tests (SETs) and integrated effect tests (IETs) are successfully performed and these results shows that SPACE/sEM code has a conservatism comparing with experimental data. Finally, plant calculations of SBLOCA for APR1400 are conducted as described below. - Break location sensitivity : DVI line, hot-leg, cold-leg, pump suction leg. - Break size spectrum : 0.4ft{sup 2}∼0.02ft{sup 2}(DVI) 0.5ft{sup 2}∼0.02ft{sup 2}(hot-leg, cold-leg, pump suction leg) This paper deals with break size spectrum analysis of cold-leg break accidents. Based on the calculation results, emergency core cooling system (ECCS) performances of APR1400 and typical SBLOCA phenomena can be evaluated. Cold-leg SBLOCA analysis for APR1400 is performed using SPACE/sEM code under harsh environment condition. SPACE/sEM code shows the typical SBLOCA behaviors and it is reasonably predicted. Although SPACE/sEM code has conservative models and correlations based on appendix K of 10 CFR 50, PCT does not exceed the requirement (1477 K). It is concluded that ECCS in APR1400 has a sufficient performance in cold-leg SBLOCA.

  8. A PC Mathcad-based computational aid for severe accident analysis and its application to a BWR small LOCA sequence

    International Nuclear Information System (INIS)

    Wu, Laung-Kuang T.; Lee, S.J.

    2004-01-01

    A PC-based Mathcad program is used to develop a computational aid for analyzing severe accident phenomena. This computational aid uses simple engineering expressions and empirical correlations to estimate key quantities and timings at various stages of accident progressions. In this paper, the computational aid is applied to analyze an early phase of a BWR small LOCA sequence. The accident phenomena analyzed include: break flow rates, boiled-up water level in the core, core uncovery time, depressurization of the reactor pressure vessel, core heat-up, onset of clad oxidation, hydrogen generation, and onset of fuel relocation. The results are compared with those obtained running the MAAP 3.0B code. This PC-based computational aid can be used to train plant personnel in understanding severe accident phenomena and to assist them in managing severe accidents. (author)

  9. Influence of liquid holdup in steam generator U-tubes on small break LOCA severity

    International Nuclear Information System (INIS)

    Leonard, M.T.; Perryman, J.L.; Johnson, G.W.

    1983-01-01

    The severity of small cold leg break loss-of-coolant accidents has been shown to be influenced by liquid holdup in steam generator U-tubes during pump suction loop seal formation in two experiments performed in the Semiscale Mod-2A facility. The core coolant level can be depressed lower than previously thought possible due to a positive hydrostatic head across the steam generators caused by delayed drainage of liquid from the upflow side of the U-tubes. The significance of a lower core coolant level depression is the potential for a more severe temperature excursion occurring during the coolant boiloff phase subsequent to loop seal clearing and prior to accumulator injection. Presented in this paper are the experimental data analysis and supporting computer code calculations that led to these conclusions

  10. Large Break LOCA Analysis with New downcomer Nodalizaion and Multi-Dimensional Model and Effect of Cross flow option in MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Hyung-wook; Lee, Sang-yong; Oh, Seung-jong; Kim, Woong-bae [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    The phenomena of LOCA have been investigated for long time. The most extensive research project for LOCA was the 2D/3D program experiments. The results of the 2D/3D experiments show flow conditions in the downcomer during end-of-blowdown were highly multi-dimensional at full-scale. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. An LBLOCA analysis for APR1400 with new downcomer input deck was conducted using KREM with MARS-KS 1.4 Version code. Analysis was processed under LBCOCA of 100% break size of cold leg case. The authors developed input deck with new downcomer nodalizaion and Multi-Dimensional downcomer model, then implemented LOCA analysis with new input decks and compared with existing analysis results. PCT from new input and multi-dimensional input deck shows similar PCT trend from original input deck. There occurred more rapid drop of PCT from new and multidimensional input deck than original input deck. PCT from new and multidimensional input deck are satisfied with PCT design limit. It can be concluded that there occurs no acceptance criteria issue even though new and multidimensional input deck are applied to LBLOCA analysis. In future study, comparative analysis with experiment results will be implemented.

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

    Directory of Open Access Journals (Sweden)

    F. Reventós

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. 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)

  14. 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

  15. Debris transport evaluation during the blow-down phase of a LOCA using computational fluid dynamics

    International Nuclear Information System (INIS)

    Park, Jong Pil; Jeong, Ji Hwan; Kim, Won Tae; Kim, Man Woong; Park, Ju Yeop

    2011-01-01

    Highlights: → We conducted CFD simulation on the spreading of the coolant in the containment after a break of the hot leg. It is used to estimate the dispersion of the debris within the containment. → It was assumed that the small and fine debris is transported by the discharge flow so that a fraction of the small and fine debris transport can be estimated based on the amount of water. → The break flow was assumed to be a homogeneous two-phase mixture without phase separation. Isenthalpic expansion of the break flow was used to specify the inlet boundary condition of the break flow. → The fraction of the small and fine debris transported to the upper part is 73%; this value is close to the value calculated using 1D lumped-parameter codes by the USNRC and the KINS, respectively, while 48% more than the value shown in the NEI 04-07. - Abstract: The performance of the emergency recirculation water sump under the influence of debris accumulation following a loss-of-coolant accident (LOCA) has long been of safety concern. Debris generation and transport during a LOCA are significantly influenced by the characteristics of the ejected coolant flow. One-dimensional analyses previously have been attempted to evaluate the debris transport during the blow-down phase but the transport evaluation still has large uncertainties. In this work, a computational fluid dynamics (CFD) analysis was utilized to evaluate small and fine debris transport during the blow-down phase of a pressurized water reactor, OPR1000. The coolant ejected from the ruptured hot-leg was assumed to expand in an isenthalpic process. The transport of small and fine debris was assumed to be dominated by water-borne transport, and the transport fractions for the upper and lower parts of the containment were quantified based on the CFD analysis. It was estimated that 73% of small and fine debris is transported to the upper part of the containment. This value is close to the values estimated by nuclear

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  18. 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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  20. CATHARE 2 analysis of the small break LOCA experiment SP-SB-03, conducted in SPES facility

    International Nuclear Information System (INIS)

    Meloni, P.

    1995-01-01

    SPES integral test facility is a scale model of a commercial three-loop PWR plant, making the simulation of a wide range of accident scenarios possible. A Small Break Loss of Coolant test was carried out in this facility in 1991 to serve as a counterpart of tests conducted on BETHSY (France), LSTF (Japan) and LOBI (EC) facilities. A post-test analysis of this test, performed with CATHARE 2 code was realized by ENEA in the framework of the co-operation ENEA-CEA on advanced reactors. This paper presents a survey of the results of the post-test calculation. (author). 5 refs, 11 figs, 3 tabs

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. Effects of high temperature ECC injection on small and large break BWR LOCA simulation tests in ROSA-III program (RUNs 940 and 941)

    International Nuclear Information System (INIS)

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

    1990-03-01

    The ROSA-III program, of which principal results are summarized in a report of JAERI 1307, conducted small and large-break loss-of-coolant experiments (RUNs 940 and 941) with high water temperature of the emergency core cooling system (ECCS) are one of the parametric study with respect to the ECCS effect on core cooling. This report presents all the experiment results of these two tests and describes additional finding with respect to the hot ECC effects on core cooling phenomena. By comparing these two tests (water temperature of 393 K) with the standard ECC tests of RUNs 922 and 926 (water temperature of 313 K), it was found that the ECC subcooling variation had a small influence on the core cooling phenomena in 5 % small break tests but had larger influence on them in 200 % break tests. The ECC subcooling effects described in the previous report are reviewed and the temperature distribution in the pressure vessel is investigated for these four tests. (author)

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

    International Nuclear Information System (INIS)

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

    1985-11-01

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

  4. PIPER-ONE: an experimental apparatus to evaluate thermal-hydraulic transients in BWRs after small breaks

    International Nuclear Information System (INIS)

    Mazzini, M.; D'Auria, F.; Vigni, P.

    1981-01-01

    This paper deals with the state of art of the research performed at the Instituto di Impianti Nucleari of Pisa University, aiming at construction of PIPER-ONE experimental facility. PIPER-ONE program is devoted to acquire direct experience on some basic phenomena, arising in BWR plants subsequently to small breaks, and on the use of the main thermal-hydraulic codes. The research has been planned taking into consideration recent trends of the studies all over the world of small LOCA thermal-hydraulics and particular needs of nuclear safety in Italy. Cost limitations and availability of some components, already installed at the Institute Laboratory, have influenced the design of the loop. The development steps of PIPER-ONE project are presented. Particularly, the overall flowsheet of the apparatus is reported. Some results of preliminary calculation, executed by RELAP4-Mod 6 code concerning both the experimental loop and the reference BWR are shown, too. A comparison with similar facilities in the world closes the paper

  5. 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)

  6. Development of technique for estimating primary cooling system break diameter in predicting nuclear emergency event sequence

    International Nuclear Information System (INIS)

    Tatebe, Yasumasa; Yoshida, Yoshitaka

    2012-01-01

    If an emergency event occurs in a nuclear power plant, appropriate action is selected and taken in accordance with the plant status, which changes from time to time, in order to prevent escalation and mitigate the event consequences. It is thus important to predict the event sequence and identify the plant behavior resulting from the action taken. In predicting the event sequence during a loss-of-coolant accident (LOCA), it is necessary to estimate break diameter. The conventional method for this estimation is time-consuming, since it involves multiple sensitivity analyses to determine the break diameter that is consistent with the plant behavior. To speed up the process of predicting the nuclear emergency event sequence, a new break diameter estimation technique that is applicable to pressurized water reactors was developed in this study. This technique enables the estimation of break diameter using the plant data sent from the safety parameter display system (SPDS), with focus on the depressurization rate in the reactor cooling system (RCS) during LOCA. The results of LOCA analysis, performed by varying the break diameter using the MAAP4 and RELAP5/MOD3.2 codes, confirmed that the RCS depressurization rate could be expressed by the log linear function of break diameter, except in the case of a small leak, in which RCS depressurization is affected by the coolant charging system and the high-pressure injection system. A correlation equation for break diameter estimation was developed from this function and tested for accuracy. Testing verified that the correlation equation could estimate break diameter accurately within an error of approximately 16%, even if the leak increases gradually, changing the plant status. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Oh, Duk Ju; Lee, Kang Moon

    1996-07-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

    Some aspects of large break LOCA analysis (steam binding, oscillatory reflooding, top-down reflooding) are expected to be improved in advanced system codes from more detailed description of flows by adding a third field for droplets. The future system code CATHARE-3 is under development by CEA and supported by EDF, AREVA-NP and IRSN in the frame of the NEPTUNE project and this paper shows some preliminary results obtained in reflooding conditions. A three-field model has been implemented, including vapor, continuous liquid and liquid droplet fields. This model features a set of nine equations of mass, momentum and energy balance. Such a model allows a more detailed description of the droplet transportation from core to steam generator, while countercurrent flow of continuous liquid is allowed. Code assessment against reflooding experiments in an isolated rod bundle mockup is presented, using 1D meshing of the bundle. Comparisons of CATHARE-3 simulations against data series from PERICLES and RBHT full scale experiments show satisfactory results. Quench front motions are well predicted, as well as clad temperatures in most of the tested runs. The BETHSY 6.7C Integral Effect Test simulating the gravity driven Reflooding process in a scaled PWR circuit is then compared to CATHARE-3 simulation. The three-field model is applied in several parts of the circuit : core, upper plenum, hot leg and steam generator, represented by either 1D or 3D modules, while the classic 6-equation model is used in the other parts of the loop. A short analysis of the results is presented. (author)

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

    International Nuclear Information System (INIS)

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

    2010-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

  12. Assessment of RELAP/MOD2 using large break loss-of-coolant experimental data

    International Nuclear Information System (INIS)

    Kao, L.; Liao, L.Y.; Liang, K.S.; Wang, S.F.; Chen, Y.B.

    1989-01-01

    In this paper assessment of RELAP5/MOD2 using LOFT L2-5 and Semiscale S-06-3 tests are performed to provide information of the code capability and its limitation in analyzing large break LOCA of a nuclear power plant. Experiments L2-5 and S-06-3 are conducted to simulate a hypothetical LOCA which results from a 200% double-ended offset shear break in the cold-leg of a typical pressurized water reactor by utilizing scaling facilities of the LOFT and Semiscale Mod-1 systems, respectively. The RELAP5/MOD2 calculations for both tests begin with break initiation and subsequent blowdown, continue through lower plenum refill, core reflood, and terminate with corewide quench. Major phenomena of both large break loss-of-coolant tests are well predicted by RELAP5/MOD2. The results indicate that the break flow and system pressure are reasonably calculated. The cladding temperature response during blowdown period, which is the major importance to a large break LOCA, calculated by RELAP5/MOD2 shows good agreement with the test data

  13. 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

  14. 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

  15. Break model comparison in different RELAP5 versions

    International Nuclear Information System (INIS)

    Parzer, I.

    2003-01-01

    The presented work focuses on the break flow prediction in RELAP5/MOD3 code, which is crucial to predict core uncovering and heatup during the Small Break Loss-of-Coolant Accidents (SB LOCA). The code prediction has been compared to the IAEA-SPE-4 experiments conducted on the PMK-2 integral test facilities in Hungary. The simulations have been performed with MOD3.2.2 Beta, MOD3.2.2 Gamma, MOD3.3 Beta and MOD3.3 frozen code version. In the present work we have compared the Ransom-Trapp and Henry-Fauske break model predictions. Additionally, both model predictions have been compared to itself, when used as the main modeling tool or when used as another code option, as so-called 'secret developmental options' on input card no.1. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  17. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  1. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

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

  2. Scaling and instrumentation of the LOFT facility

    International Nuclear Information System (INIS)

    Modro, S.M.; Goodrich, L.D.; McPherson, G.D.

    1985-01-01

    This paper describes the LOFT experimental facility and instrumentation of the facility during small break loss-of-coolant experiments. Basic scaling considerations applied in the facility design are presented. Because LOFT was not designed with emphasis on small break LOCA some atypicalities with regard to small break transients are discussed. Review of important small break LOCA phenomena observed during the experiments and their measurability is provided

  3. Application of realistic (best- estimate) methodologies for large break loss of coolant (LOCA) safety analysis: licensing of Westinghouse ASTRUM evaluation model in Spain

    International Nuclear Information System (INIS)

    Lage, Carlos; Frepoli, Cesare

    2010-01-01

    When the LOCA Final Acceptance Criteria for Light Water Reactors was issued in Appendix K of 10CFR50 both the USNRC and the industry recognized that the rule was highly conservative. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the USNRC began a research program to identify the degree of conservatism in those models permitted in the Appendix K rule and to develop improved thermal-hydraulic computer codes so that realistic accident analysis calculations could be performed. The overall results of this research program quantified the conservatism in the Appendix K rule and confirmed that some relaxation of the rule can be made without a loss in safety to the public. Also, from a risk-informed perspective it is recognized that conservatism is not always a complete defense for lack of sophistication in models. In 1988, as a result of the improved understanding of LOCA phenomena, the USNRC staff amended the requirements of 10 CFR 50.46 and Appendix K, 'ECCS Evaluation Models', so that a realistic evaluation model may be used to analyze the performance of the ECCS during a hypothetical LOCA. Under the amended rules, best-estimate plus uncertainty (BEPU) thermal-hydraulic analysis may be used in place of the overly prescriptive set of models mandated by Appendix K rule. Further guidance for the use of best-estimate codes was provided in Regulatory Guide 1.157 To demonstrate use of the revised ECCS rule, the USNRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology as an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis. More recently the CSAU principles have been generalized in the Evaluation Model Development and Assessment Process (EMDAP) of Regulatory Guide 1.203. ASTRUM is the Westinghouse Best Estimate Large Break LOCA evaluation model applicable to two-, three

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

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

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

  5. Upper plenum break LOCA investigation in the ISB-VVER and PSB-VVER facilities

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V. N.; Melikhov, O. I.; Lipatov, I. A.; Nikonov, S. M.; Dremin, G. I.; Galchanskaya, S. A.; Gashenko, M. P.; Rovnov, A. A.; Kapustin, A. V.; Elkin, I. V. [EREC, Moscow (Russian Federation)

    2003-07-01

    The capability to define the actual NPP transient/accident scenario depends to a great extent on facilities' scaling and reliability of the system thermalhydraulic codes which, in turn, are assessed against the experimental data taken in the same facilities. At the present time, it is received fact that the rigorous modeling of the cumulative set of all thermalhydraulic processes in the plant primary and secondary sides during accident is unfeasible. Therefore, the extrapolation of the facilities loops behavior to the actual systems constitutes a fundamental problem in this area. In the paper, some aspects for the problem have been discussed in the course of comparative analysis of the data derived from the 11 % upper plenum break LOCA tests performed in the PSB-VVER and ISB-VVER integral test facilities under the close scenarios. Both facilities, PSB-VVER and ISB-VVER, are modeled the same VVER-1000 reactor in different scales. The thermalhydraulic behavior of the primary systems in both facilities has been discussed in the paper, and shown to be similar. Also, the attention has been focused upon the discrepancies in the significant variables trends. The discrepancies are shown to be caused by influence of peculiarities of the facilities hardware and due to the scale factor. The scaling study is an important aspect of the thermalhydraulic codes verification procedure. Being qualified against the experimentally simulated accident sequence in two test facilities of different scales, the thermalhydraulic codes will be capable of evaluation of the prototype behavior to greater accuracy.

  6. 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

  7. 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)

  8. Verification of LOCA/ECCS analysis codes ALARM-B2 and THYDE-B1 by comparison with RELAP4/MOD6/U4/J3

    International Nuclear Information System (INIS)

    Shimizu, Takashi

    1982-08-01

    For a verification study of ALARM-B2 code and THYDE-B1 code which are the component of the JAERI code system for evaluation of BWR ECCS performance, calculations for typical small and large break LOCA in BWR were done, and compared with those by RELAP4/MOD6/U4/J3 code. This report describes the influences of differences between the analytical models incorporated in the individual code and the problems identified by this verification study. (author)

  9. Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2009-01-01

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  11. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  13. 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.

  14. 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

  15. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

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

    1991-10-01

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

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

    International Nuclear Information System (INIS)

    Sabourin, G.; Huynh, H.M.

    1997-01-01

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

  17. 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)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Steam leak detection in advance reactors via acoustics method

    International Nuclear Information System (INIS)

    Singh, Raj Kumar; Rao, A. Rama

    2011-01-01

    Highlights: → Steam leak detection system is developed to detect any leak inside the reactor vault. → The technique uses leak noise frequency spectrum for leak detection. → Testing of system and method to locate the leak is also developed and discussed in present paper. - Abstract: Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.

  20. Recirculation pump suction line 2.8% break integral test at ROSA-III with HPCS failure, RUN 984

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Anoda, Yoshinari; Tasaka, Kanji; Kumamaru, Hiroshige; Nakamura, Hideo; Yonomoto, Taisuke; Murata, Hideo; Shiba, Masayoshi

    1984-06-01

    This report presents the experimental data of 2.8% suction line break test RUN 984 at ROSA-III, which was conducted as one of counterpart tests to FIST program sponsored by GE, EPRI and USNRC. The similarity study between the ROSA-III and FIST tests is on the way. The report also presents the information on the ROSA-III test facility, experiment results and the effects of the ADS flow rate and the MSIV trip level comparing with the previously conducted ROSA-III small break tests, RUNs 920 and 922. Major conclusions obtained are as follows. (1) Change of the MSIV trip level from L2 to L1 gives delay of MSIV closure and longer actuation of pressure control system in a small break LOCA. (2) Larger ADS flow gives faster depressurization rate and earlier ECCS actuation, which results in shorter fuel rod dryout period and lower PCT. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  2. 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

  3. Assessment of RELAP/MOD3 using BETHSY 6.2TC 6-inch cold leg side break comparative test

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Jeong, Jae-Jun; Chang, Won-Pyo; Kim, Dong-Su

    1996-10-01

    This report presents the results of the RELAP5/MOD3 Version 7j assessment on BETHSY 6.2TC. BETHSY 6.2TC test corresponding to a six inch cold leg break LOCA of the Pressurizer Water Reactor(PWR). The primary objective of the test was to provide reference data of two facilities of different scales (BETHSY and LSTF facility). On the other hand, the present calculation aims at analysis of RELAP5/N4OD3 capability on the small break LOCA simulation, The results of calculation have shown that the RELAP5/MOD3 reasonably predicts occurrences as well as trends of the major phenomena such as primary pressure, timing of loop seal clearing, liquid hold up, etc. However, some disagreements also have been found in the predictions of loop seal clearing, collapsed core water level after loop seal clearing, and accumulator injection behaviors. For better understanding of discrepancies in same predictions, several sensitivity calculations have been performed as well. These include the changes of two-phase discharge coefficient at the break junction and some corrections of the interphase drag term. As result, change of a single parameter has not improved the overall predictions and it has been found that the interphase drag model has still large uncertainties

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  5. TRAC-PD2 modeling of LOFT and PWR small cold-leg breaks

    International Nuclear Information System (INIS)

    Knight, T.D.; Willcutt, G.J.E. Jr.; Lime, J.F.

    1981-01-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light-water reactors. TRAC-PD2, the latest publicly released version of the code, is currently being tested against small-break and other transients in experimental facilities; it is also being used to analyze postulated accidents in commercial power reactors. Calculated results for LOFT small-break experiments are compared to data, and the results from two small-break calculations for two different reactor systems are presented. It is concluded that TRAC-PD2 is useful for the analysis of cold-leg small-break accidents

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  7. FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE

    International Nuclear Information System (INIS)

    1993-01-01

    1 - Description of test facility: BWR/6-218 standard plant. A full size bundle with electrically heated rods is used to simulate the reactor core. A scaling ratio of 1/624 is applied in the design of the system components. Key features of the FIST facility include: (1) Full height test vessel and internals; (2) correctly scaled fluid volume distribution; (3) simulation of ECCS, S/RV, and ADS; (4) level trip capability; (5) heated feedwater supply system, which provides the capability for steady state operation. 2 - Description of test: Test 6IB1 investigates system responses to an intermediate break in the recirculation suction line. BWR system licensing evaluations for various size recirculation break LOCA's indicates that a break size of about 0.2 sq.ft., without LPCS operation, is the highest PCT case for the intermediate break LOCA. Test 6IB1 simulates this event

  8. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  11. Experimental study on fundamental phenomena in HTGR small break air-ingress accident

    International Nuclear Information System (INIS)

    Kim, Jae Soon; Hwang, Jin-Seok; Kim, Eung Soo; Kim, Byung Jun; Oh, Chang Ho

    2016-01-01

    Highlights: • Air-ingress phenomena on the small break in a HTGR are experimentally investigated. • Experiment is investigated for various break sizes, angles, and density ratios. • Maximum air-ingress rate is observed at 120° in break angle. • This study reveals that air-ingress in the small break is governed by; buoyancy and flow inertia. • A non-dimensional parameter is newly proposed to determine the air-ingress flow regimes. • Newly proposed parameter is based on buoyancy versus inertia force. - Abstract: This study experimentally investigates fundamental phenomena in the HTGR small break air-ingress accident. Several important parameters including density ratio, break angle, break size, and main flow velocity are considered in the measurement and the analysis. The test-section is made of a circular pipe with small holes drilled around the surface and it is installed in the helium/air flow circulation loop. Oxygen concentrations and flow rates are recorded during the tests with fixed break angles, break sizes, and flow velocities for measurement of the air-ingress rates. According to the experimental results, the higher density difference leads to the higher rates of air-ingress with large sensitivity of the break angles. It is also found that the break angle significantly affects the air-ingress rates, which is gradually increased from 0° to 120° and suddenly decreased to 180°. The minimum air ingress rate is found at 0° and the maximum, at 110°. The air-ingress rate increases with the break size due to the increased flow-exchange area. However, it is not directly proportional to the break area due to the complexity of the phenomena. The increased flow velocity in the channel inside enhances the air-ingress process. However, among all the parameters, the main flow velocity exhibits the lowest effect on this process. In this study, the Froude Number relevant to the small break air-ingress conditions are newly defined considering both heavy

  12. Post accidental small breaks analysis

    International Nuclear Information System (INIS)

    Depond, G.; Gandrille, J.

    1980-04-01

    EDF ordered to FRAMATOME by 1977 to complete post accidental long term studies on 'First Contrat-Programme' reactors, in order to demonstrate the safety criteria long term compliance, to get information on NSSS behaviour and to improve the post accidental procedures. Convenient analytical models were needed and EDF and FRAMATOME respectively developped the AXEL and FRARELAP codes. The main results of these studies is that for the smallest breaks, it is possible to manually undertake cooling and pressure reducing actions by dumping the steam generators secondary side in order to meet the RHR operating specifications and perform long term cooling through this system. A specific small breaks procedure was written on this basis. The EDF and FRAMATOME codes are continuously improved; the results of a French set of separate effects experiments will be incorporated as well as integral system verification

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  15. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    Yano, Toshikazu; Miyazaki, Noriyuki; Isozaki, Toshikuni

    1982-10-01

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

  16. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    International Nuclear Information System (INIS)

    Beghini, M.; D'Auria, F.; Galassi, G.M.; Vitale, E.

    1997-01-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs

  17. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    Energy Technology Data Exchange (ETDEWEB)

    Beghini, M; D` Auria, F; Galassi, G M; Vitale, E [Universita degli Studi di Pisa, Dipt. di Costruzioni Meccaniche e Nucleari, Pisa (Italy)

    1997-09-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs.

  18. 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.)

  19. Applicability of TASS/SMR using drift flux model for SMART LOCA analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young-Jong, E-mail: chung@kaeri.re.kr; Kim, Soo Hyung; Lee, Gyu Hyeung; Lee, Won Jae

    2013-09-15

    Highlights: • SMART plant and TASS/SMR code have been developed by KAERI. • TASS/SMR code adopts a drift flux model to consider relative velocity under two-phase condition. • Drift flux model in TASS/SMR is validated using separate effect test results. • Applicability of TASS/SMR using drift flux model for SMART LOCA analysis is confirmed. -- Abstract: Small reactors can apply to local power demands or remote areas. SMART, which can produce 90 MWe of electricity and 40,000 tons/day of sea-water desalination for a 100,000 population city, is a promising advanced integral type small reactor. The thermal hydraulic analysis code with a drift flux model, TASS/SMR, was developed for a conservative simulation of a small break loss of coolant accident in SMART. Taking into account SMART-specific inherent characteristics, the code adopts the Chexal–Lellouche correlation for a drift flux model. The capability of TASS/SMR code is validated using the results of the experimental data. The code predicts conservatively the void distribution compared with the experimental data. TASS/SMR calculation predicts reasonably major phenomena for the SBLOCA and is more conservative than or nearly the same as the results of the best estimated realistic system code, MARS. TASS/SMR code can be used for a SBLOCA analysis of SMART.

  20. Applicability of TASS/SMR using drift flux model for SMART LOCA analysis

    International Nuclear Information System (INIS)

    Chung, Young-Jong; Kim, Soo Hyung; Lee, Gyu Hyeung; Lee, Won Jae

    2013-01-01

    Highlights: • SMART plant and TASS/SMR code have been developed by KAERI. • TASS/SMR code adopts a drift flux model to consider relative velocity under two-phase condition. • Drift flux model in TASS/SMR is validated using separate effect test results. • Applicability of TASS/SMR using drift flux model for SMART LOCA analysis is confirmed. -- Abstract: Small reactors can apply to local power demands or remote areas. SMART, which can produce 90 MWe of electricity and 40,000 tons/day of sea-water desalination for a 100,000 population city, is a promising advanced integral type small reactor. The thermal hydraulic analysis code with a drift flux model, TASS/SMR, was developed for a conservative simulation of a small break loss of coolant accident in SMART. Taking into account SMART-specific inherent characteristics, the code adopts the Chexal–Lellouche correlation for a drift flux model. The capability of TASS/SMR code is validated using the results of the experimental data. The code predicts conservatively the void distribution compared with the experimental data. TASS/SMR calculation predicts reasonably major phenomena for the SBLOCA and is more conservative than or nearly the same as the results of the best estimated realistic system code, MARS. TASS/SMR code can be used for a SBLOCA analysis of SMART

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

    International Nuclear Information System (INIS)

    Winkler, F.; Krebs, W.D.

    1989-01-01

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

  2. BWR full integral simulation test (FIST) pretest predictions with TRACBO2

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.

    1984-01-01

    The Full Integral Simulation Test program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An analytical method development program is underway to extend the BWR-TRAC computer code to model reactor kinetics and major interfacing systems, including balance-of-plant, to improve application modeling flexibility, and to reduce computer running time. An experimental program is underway in a new single bundle system test facility to extend the large break loss-of-coolant accident LOCA data base to small breaks and operational transients. And a method qualification program is underway to test TRACBO2 against experiments in the FIST facility. The recently completed Phase 1 period included a series of LOCA and power transient tests, and successful pretest analysis of the large and small break LOCA tests with TRACBO2. These comparisons demonstrate BWR-TRAC capability for small and large break analysis, and provide detailed understanding of the phenomena

  3. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm2, simulated with RELAP5 code

    International Nuclear Information System (INIS)

    Borges, Eduardo M.; Sabundjian, Gaiane

    2015-01-01

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm 2 -rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

  4. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujita, R.K.

    1985-01-01

    A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer

  5. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujits, R.K.

    1985-01-01

    A computer code (TRAC-PFI/MODI; denoted as TRAC) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the Once-Through Integral Systems (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and saturation, intermittent reactor coolant system circulation, boiler-condenser mode and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool- and auxiliary- feedwater initiated boiler-condenser mode heat transfer

  6. Small break critical discharge: The roles of vapor and liquid entrainment in a stratified two-phase region upstream of the break

    International Nuclear Information System (INIS)

    Schrock, V.E.; Revankar, S.T.; Mannheimer, R.; Wang, C.H.

    1986-12-01

    The main objective of this research program was to perform an experimental investigation on the phenomena of two-phase critical flow through small break from a horizontal pipe which contained a stratified two phase flow. Stagnation conditions investigated were saturated steam-water, and air-cold water at pressures ranging from 0.37 MPa to 1.07 MPa. The small breaks employed were cylindrical tubes of diameters 3.96 mm, 6.32 mm, and 10.1 mm with sharp-edged entrance. For breaks resulting from a small hole in a primary coolant pipe or in a small pipe, a sharp-edged orifice or a sharp-edged tube can be the approximation

  7. 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

  8. 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

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

    International Nuclear Information System (INIS)

    Kim, Jong Seog

    2013-01-01

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

  10. Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  12. Design basis neutronics calculations for NRU-LOCA experiments

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  13. About the use of the CATHARE code for best estimate Large Break LOCA calculations: benefits for Safety and constraints

    International Nuclear Information System (INIS)

    Vacher, J.L.

    1994-01-01

    Since 1979, EDF has participated to the development of CATHARE, a best estimate accidental thermalhydraulic code, in collaboration with CEA and FRAMATOME. EDF is now investigating the use of this code for licensing studies and particularly for Large Break LOCA calculations. Until now, the work done at EDF, in relation with FRAMATOME and CEA, has mainly focused on the physical analysis of the transient and on the identification of the key phenomena. This task is a necessary step before uncertainty evaluation. To illustrate this point of view, a peculiar example of calculated Large Break transients for a 900 MW three loop plant is presented. In one of these calculations, a high value of Peak Cladding Temperature was obtained. This peculiar scenario was initiated by a large entrainment of water to the steam generators at the very beginning of the reflooding stage, followed by a strong pressurization which led to a lasting draining of the reactor vessel. The physical phenomena which determine the existence and amplitude of this scenario were identified and their influence was explained: condensation at the accumulator injection, heat exchange in the core, entrainment process to the steam generators. It appeared obvious that the large observed uncertainty was associated to only a few parameters. Although this peculiar system behaviour was obtained for only a particular combination of parameters and a narrow range of thermalhydraulic conditions, the capability of the code to simulate these phenomena was investigated in regard to experimental data. It was concluded that this scenario was definitely unrealistic on a reactor. Nevertheless, this peculiar example tends to demonstrate, firstly, that the use of a best-estimate code improves Safety as it makes possible to point out physical phenomena that could not be considered when using non mechanistic codes, secondly, that the uncertainty evaluation must be guided by a pertinent physical analysis of the transient, focusing

  14. Shut-down conditions, emergency cooling and essential services

    International Nuclear Information System (INIS)

    Belda, W.

    1977-01-01

    1) Introduction: Summary of system technology and reactor protection equipment. 2) Definitions. 3) LOCA: a) blowdown and refilling phase; b) jet and reaction forces; c) flow and heat transfer behavior in the core; d) behavior of the heater rods; e) core melting. 4) Protection against and during LOCA: a) general measures; b) break of a primary coolant pipe; c) break of a small pipe; d) break of a secondary pipe. (orig.) [de

  15. 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

  16. 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

  17. 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

  18. Scaling effects concerning the analysis of small break experiments

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.

    1985-01-01

    Some scaling effects related to the experimental facilities as well as to the analytical models used for the design and safety analysis of nuclear power plants are discussed or the basis of phenomena expected to occur during small-break loss - of - coolant accidents. The results of isolated small-break experiments should not be directly extrapolated to the safety analysis of commercial reactors, due to the scaling distortions inherent to the test facilities. With respect to the analytical models used to simulate thermohydraulic processes in experimental facilities, their eventual dependence relative to the system dimension should be examined in order to assess their applicability to the safety analysis of commercial power plants. (Author) [pt

  19. Accidents and transients analyses of a super fast reactor with single flow pass core

    International Nuclear Information System (INIS)

    Sutanto,; Oka, Yoshiaki

    2014-01-01

    Highlights: • Safety analysis of a Super FR with single flow pass core is conducted. • Loss of feed water flow leads to a direct effect on the loss of fuel channel flow. • The core pressure is sensitive to LOCA accidents due to the direct effect. • Small LOCA introduces a critical break. • The safety criteria for all selected events are satisfied. - Abstract: The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST

  20. A 25% double-ended LOCA in the PSB-VVER facility

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Lipatov, I.A.; Dremin, G.I.

    2003-01-01

    The paper presents the results of the experimental investigation in the PSB-VVER facility and post-test analysis with RELAP5/MOD3.3 of the thermal-hydraulic response of the PSB-VVER to a 25% double-ended Hot Leg Break (HLB). The test scenario included loss of off-site power concurrently with the scram-signal and the safety system operation as described in the reference VVER-1000 operational manual in the case of this type of accident assuming one diesel-generator failure. The key transient parameter trends as well as sequence of events and phenomena are given in the paper. RELAP5/MOD3.3 a post-test analysis has been performed using the experimental data gained as a base. The reasonable qualitative agreement between the key calculated and measured variables has been shown. The quantitative code accuracy evaluation has shown that the total average amplitude of the main parameters' deviations AA tot tot < 0.28 that corresponds to satisfactory quality of the VVER-1000 hot leg guillotine break LOCA modeling in the PSB-VVER. (author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  2. FIST small break accident analysis with BWR TRACBO2-pretest predictions

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    The BWR Full Integral Simulation Test (FIST) program includes experimental simulation and analytical evaluation of BWR thermal-hydraulic phenomena during transient events. One such event is a small size break in the suction line of one of the recirculation pumps. The results from a test simulating this transient in the FIST facility are compared with a system analysis using the Transient Reactor Analysis Code (TRACB02). This comparison demonstrates BWR-TRAC capability for small break analyses and provides detailed understanding of the phenomena

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

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

    International Nuclear Information System (INIS)

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

    1996-09-01

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

  5. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

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

    International Nuclear Information System (INIS)

    Griggs, D.P.

    1993-03-01

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

  7. Intermediate-break LOCA analyses for the AP600 design

    International Nuclear Information System (INIS)

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

    1995-01-01

    A postulated double-ended guillotine break of a direct-vessel-injection line in an AP600 plant has been analyzed. This event is characterized as an intermediate break loss-of-coolant accident (IBLOCA). Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PFl/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. The key processes occurring in an AP600 during a IBLOCA are primary coolant system depressurization, inventory depletion, inventory replacement via emergency core coolant injection, continuous core cooling, and long-term decay heat rejection to the atmosphere. 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 latter phase of the long-term cooling period, the interval when sump injection and containment cooling processes are important

  8. 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)

  9. 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)

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

    International Nuclear Information System (INIS)

    Lee, Sang Il

    1992-02-01

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

  11. 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

  12. 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)

  13. Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

    Directory of Open Access Journals (Sweden)

    Eltayeb Yousif

    2017-01-01

    Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

  14. CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

    Directory of Open Access Journals (Sweden)

    Luben Sabotinov

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-10-01

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

  16. Experimental analysis of the power curve sensitivity test series at ROSA-III

    International Nuclear Information System (INIS)

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

    1985-01-01

    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 LOCA and ECCS tests. Seven recirculation pump suction line break LOCA experiments were conducted at the ROSA-III facility in order to examine the effect of the initial stored heat of a fuel rod on the peak cladding temperature (PCT). The break size was changed from 200% to 5% in the test series and a failure of a high pressure core spray (HPCS) diesel generator was assumed. Three power curves which represented conservative, realistic and zero initial stored heat, respectively, were used. In a large break LOCA such as 200% or 50% breaks, the initial stored heat in a fuel rod has a large effect on the cladding surface temperature because core uncovery occurs before all the initial stored heat is released, whereas in a small break LOCA such as a 5% break little effect is observed because core uncovery occurs after the initial stored heat is released. The maximum PCTs for the conservative initial stored heat case was 925 K, obtained in the 50% break experiment, and that for the realistic initial stored heat case was 835 K, obtained in the 5% break experiment. (orig./HP)

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Analysis of factors affecting the LOCA test quality

    International Nuclear Information System (INIS)

    Wang Lu

    2008-01-01

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

  19. Improved Methodology of MSLB M/E Release Analysis for OPR1000

    International Nuclear Information System (INIS)

    Park, Seok Jeong; Kim, Cheol Woo; Seo, Jong Tae

    2006-01-01

    A new mass and energy (M/E) release analysis methodology for the equipment environmental qualification (EEQ) on loss-of-coolant accident (LOCA) has been recently developed and adopted on small break LOCA EEQ. The new methodology for the M/E release analysis is extended to the M/E release analysis for the containment design for large break LOCA and the main steam line break (MSLB) accident, and named KIMERA (KOPEC Improved Mass and Energy Release Analysis) methodology. The computer code systems used in this methodology is RELAP5K/CONTEMPT4 (or RELAP5-ME) which couples RELAP5/MOD3.1/K with enhanced M/E model and LOCA long term model, and CONTEMPT4/ MOD5. This KIMERA methodology is applied to the MSLB M/E release analysis to evaluate the validation of KIMERA methodology for MSLB in containment design. The results are compared with the OPR 1000 FSAR

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  1. Assessment of TRAC-BD1/MOD1 using FIST data

    International Nuclear Information System (INIS)

    Jo, J.H.; Connell, H.R.

    1985-01-01

    This report is concerned with the assessment of the TRAC-BD1/MOD1 Code, developed at Idaho National Engineering Laboratory. The assessment was conducted using data from the FIST (Full Integral Simulation Test) facility, which is a BWR safety test facility which was built to investigate small break LOCA and operational transients in BWR's and to complement earlier large break LOCA test results from TLTA (Two-Loop Test Apparatus). 21 figs

  2. A comparison of the effect of the first and second upwind schemes on the predictions of the modified RELAP5/MOD3

    Energy Technology Data Exchange (ETDEWEB)

    Analytis, G.Th. [Paul Scherrer Institute (PSI), Villigen (Switzerland)

    1995-09-01

    As is well-known, both TRAC-BF1 and TRAC-PF are using the first upwind scheme when finite-differencing the phasic momentum equations. In contrast, RELAP5 uses the second upwind which is less diffusive. In this work, we shall assess the differences between the two schemes with our modified version of RELAP5/MOD3 by analyzing some transients of interest. These will include the LOFT LP-LB-1 and LOBI small break LOCA (SB-LOCA) BL34 tests, and a commercial PWR 200% hypothetical large break LOCA (LB-LOCA). In particular, we shall show that for some of these transients, the employment of the first upwind scheme results in significantly different code predictions than the ones obtained when the second upwind scheme is used.

  3. Symmetry breaking in small rotating clouds of trapped ultracold Bose atoms

    International Nuclear Information System (INIS)

    Dagnino, D.; Barberan, N.; Riera, A.; Osterloh, K.; Lewenstein, M.

    2007-01-01

    We study the signatures of rotational and phase symmetry breaking in small rotating clouds of trapped ultracold Bose atoms by looking at rigorously defined condensate wave function. Rotational symmetry breaking occurs in narrow frequency windows, where energy degeneracy between the lowest energy states of different total angular momentum takes place. This leads to a complex condensate wave function that exhibits vortices clearly seen as holes in the density, as well as characteristic local phase patterns, reflecting the appearance of vorticities. Phase symmetry (or gauge symmetry) breaking, on the other hand, is clearly manifested in the interference of two independent rotating clouds

  4. An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-03-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

  5. Plans and status of RELAP5/MOD3

    International Nuclear Information System (INIS)

    Weaver, W.L.

    1989-01-01

    RELAP5/MOD3 is a pressurized water reactor (PWR) system analysis code being developed jointly by the US Nuclear Regulatory Commission (USNRC) and consisting of several of the countries that are members of the International Code Assessment and Applications Program (ICAP). This code development program is called the ICAP Code Improvement Program. The mission of the RELAP5/MOD3 code improvement program is to develop a code version suitable for the analysis of all transients and postulated accidents in PER systems including both large and small break loss of coolant accidents (LOCA's) as well as the full range of operational transients. The emphasis of the RELAP5/MOD3 development will be on large break LOCA since previous versions of RELAP5 were developed for and assessed against small break LOCA and operation transient test data. The paper discusses the various code models to be improved and presents the results of work completed to date

  6. Experiment data of 200% recirculation pump discharge line break integral test run 961 with HPCS failure at ROSA-III and comparison with results of suction line break tests

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Tasaka, Kanji; Nakamura, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Murata, Hideo; Yonomoto, Taisuke; Shiba, Masayoshi

    1984-03-01

    This report presents the experimental data of RUN 961, a 200% double-ended break test at the recirculation pump discharge line in the ROSA-III test facility. The ROSA-III test facility is a volumetrically scaled (1/424) system of the BWR/6. The facility has the electrically heated core, the break simulator and the scaled ECCS (Emergency Core Cooling System). The MSIV (Main Steam Isolation Valve) closure and the ECCS actuation were tripped by the liquid level in the upper downcomer. The PCT (Peak Cladding Temperature) was 894 K, which was 107 K higher than a 200% pump suction line break test (RUN 926) due to the smaller depressurization rate. The effect of break location on transient LOCA phenomena was clarified by comparing the discharge and suction break tests. The whole core was quenched 71 s after LPCI actuation and the effectiveness of ECCS has been confirmed. (author)

  7. 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)

  8. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. 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

  11. 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.

  12. Audit calculation for the LOCA methodology for KSNP

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-11-15

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

  13. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Ohtsu, Iwao [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokaimura (Japan)

    2017-08-15

    An experiment using the Primaerkreislaeufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  14. ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

    Directory of Open Access Journals (Sweden)

    Takeshi Takeda

    2017-08-01

    Full Text Available An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF on a cold leg small-break loss-of-coolant accident with an accident management (AM measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

  15. LOFT/LP-02-6, Loss of Fluid Test, 1. OECD Large Break 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 LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The fourth OECD LOFT experiment was conducted on 3 October 1983. This was the first OECD LOFT large break experiment. The initial and boundary conditions were chosen to be representative of USNRC licensing limits for commercial PWRs. This included loss of off-site power coincident with LOCA initiation. This experiment included the first use in LOFT of pressurized fuel rods in the center bundle. The experiment was initiated by opening the quick-opening blow-down valves in the broken 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

  16. Special small-break applications with TRAC

    International Nuclear Information System (INIS)

    Dobranich, D.; DeMuth, N.S.; Henninger, R.J.; Burns, R.D. III.

    1981-01-01

    Input models for the Transient Reactor Analysis Code (TRAC) are described and applications of these models to reactor transients involving small breaks in the primary coolant pressure boundary are demonstrated. The operation of the primary overpressure protection system (relief and safety valves) and the thermal-hydraulic response of the reactor to these transients are obtained from numerical simulations. Also, the effects of steam generator recirculation, steam generator tube rupture, Emergency Core Cooling (ECC) injection and reactivity feedback on the course and consequences of these transients are investigated. These models allow reliable predictions of accident signatures that can help determine the adequacy of equipment and procedures at nuclear power plants to prevent and to control severe accidents

  17. An IPSN research programme to resolve pending LOCA issues

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  19. A study on design improvement of the emergency core cooling system for a nuclear ship reactor

    International Nuclear Information System (INIS)

    Naruko, Yoshinori

    1982-01-01

    ECCS performances are predicted for the N.S. ''MUTSU'' Reactor using new computing techniques. The actual performances are found to be inadequate with regard to the total injection flow rate and the infection head of the High Pressure Injection System (HPIS). The actual Safety Injection (SI) Signal is also shown to be inoperative in a gas phase LOCA. Because of this fact, the ECCS has been improved by replacing the existing HPIS with two large high pressure pumps and by adding ''Reactor Pressure Low-Low'' and ''Containment Pressure High'' signals. This report deals with a numerical study of the dynamical behavior of the N.S. MUTSU Reactor in the postulated small break LOCA, which has been calculated to verify the design improvement by evaluating the new ECCS performances on the basis of RELAP4/MOD6/SUS and TOODEE2/SUS codes. In conclusion, the improved system performs satisfactorily in the whole range of gas phase breaks and in the small break range of liquid phase LOCA. (author)

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

    Science.gov (United States)

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

    2018-04-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  2. 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)

  3. Progress in realistic LOCA analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  6. Advanced LOCA code uncertainty assessment

    International Nuclear Information System (INIS)

    Wickett, A.J.; Neill, A.P.

    1990-11-01

    This report describes a pilot study that identified, quantified and combined uncertainties for the LOBI BL-02 3% small break test. A ''dials'' version of TRAC-PF1/MOD1, called TRAC-F, was used. (author)

  7. Comparisons of ROSA-III and FIST BWR loss of coolant accident simulation tests

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Suzuki, Mitsuhiro; Koizumi, Yasuo

    1985-10-01

    A common understanding and interpretation of BWR system response and the controlling phenomena in LOCA transients has been achieved through the evaluation and comparison of counterpart tests performed in the ROSA-III and FIST test facilities. These facilities, which are designed to simulate the thermal-hydraulic response of BWR systems, are operated respectively by the Japan Atomic Energy Research Institute (JAERI) and the General Electric Company. Comparison is made between three types of counterpart tests, each performed under similar tests conditions in the two facilities. They are large break, small break, and steamline break LOCA's. The system responses to these tests in each facility are quite similar. The sequence of events are similar, and the timing of these events are similar. Differences that do occur are due to minor differences in modeling objectives, facility scaling, and test conditions. Parallel channel flow interactions effects in the ROSA-III four channel (half length) core, although noticeable in the large break test, do not result in major differences with the single channel response in FIST. In the small break tests the timing of events is offset by the earlier ADS actuation in FIST. The steamline test responses are similar except there is no heatup in FIST, resulting from a different ECCS trip modeling. Overall comparisons between ROSA-III and FIST system responses in LOCA tests is very good. (author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Vitanza, C.; Hrehor, M.

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. 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.

  13. Detailed CATHENA Model of the Wolsong 1 Pressure and Inventory Control System

    Energy Technology Data Exchange (ETDEWEB)

    Cha, K.H. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    The Detailed CATHENA model of Wolsong 1 is development to be able to simulate a theramal hydraulic behavior of heat transport system(HTS) Pressure and Inventory Control System(PNIC) at any power operation condition and during transient events such as mall LOCA(small loss of coolant inventory and small breaks in the primary system piping) and non-LOCA(loss of reactivity regulation, loss of flow, loss if Class IV power, loss of PNIC). (author). 12 refs., 7 figs., 6 tabs.

  14. Development of a qualified nodalization for small-break LOCA transient analysis in PSB-VVER integral test facility by RELAP5 system code

    Energy Technology Data Exchange (ETDEWEB)

    Shahedi, S. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Jafari, J., E-mail: jalil_jafari@yahoo.co [Reactors and Accelerators R and D School, Nuclear Science and Technology Research Institute, North Kargar Street, Tehran (Iran, Islamic Republic of); Boroushaki, M. [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); D' Auria, F. [DIMNP, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    This paper deals with development and qualification of a nodalization for modeling of the PSB-VVER integral test facility (ITF) by RELAP5/MOD3.2 code and prediction of its primary and secondary systems behaviors at steady state and transient conditions. The PSB-VVER is a full-height, 1/300 volume and power scale representation of a VVER-1000 NPP. A RELAP5 nodalization has been developed for PSB-VVER modeling and a nodalization qualification process has been applied for the developed nodalization at steady state and transient levels and a qualified nodalization has been proposed for modeling of the PSB ITF. The 11% small-break loss-of-coolant-accident (SBLOCA), i.e. rupture of one of the hydroaccumulators (HA) injection lines in the upper plenum (UP) region of reactor pressure vessel (RPV) below the hot legs (HL), inlets has been considered for nodalization qualification process. The influence of the different steam generator (SG) nodalizations on the RELAP5 results and on the nodalization qualification process has been examined. The 'steady state' qualification level includes checking the correctness of the initial and boundary conditions and geometrical fidelity. In the 'transient' qualification level, the time dependent results of the code calculation are compared with the experimental time trends from both the qualitative and quantitative point of view. For quantitative assessment of the results, a Fast Fourier Transform Based Method (FFTBM) has been used. The FFTBM was used to establish a range in which the steam generators nodalizations can vary.

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

    International Nuclear Information System (INIS)

    Buckalew, W.H.

    1989-07-01

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

  16. A simple blowdown code for SUPER-SARA loop conditions

    International Nuclear Information System (INIS)

    Fritz, G.

    1981-01-01

    The Super Sara test programme (SSTP) is aimed to study in pile the fuel and cluster behaviour under two types of accident conditions: - the ''Large break loss of coolant'' condition (LB-Loca), - the ''Severe fuel damage'' (SFD) in a boildown caused by a small break. BIVOL was developed for the LB-Loca situation. This code is made for a loop where essentially two volumes define the thermohydraulics during the blowdown. In the SUPERSARA loop these two volumes are represented by the hot leg and cold leg pipings together with the respective upper and lower plenum of the test section

  17. Application of code scaling, applicability and uncertainty methodology to large break LOCA analysis of two loop PWR

    International Nuclear Information System (INIS)

    Mavko, B.; Stritar, A.; Prosek, A.

    1993-01-01

    In NED 119, No. 1 (May 1990) a series of six papers published by a Technical Program Group presented a new methodology for the safety evaluation of emergency core cooling systems in nuclear power plants. This paper describes the application of that new methodology to the LB LOCA analysis of the two loop Westinghouse power plant. Results of the original work were used wherever possible, so that the analysis was finished in less than one man year of work. Steam generator plugging level and safety injection flow rate were used as additional uncertainty parameters, which had not been used in the original work. The computer code RELAP5/MOD2 was used. Response surface was generated by the regression analysis and by the artificial neural network like Optimal Statistical Estimator method. Results were compared also to the analytical calculation. (orig.)

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

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

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

  19. TRAC-PF1/MOD2 best-estimate analysis of a large-break LOCA in a 15 x 15 generic four-loop Westinghouse nuclear power plant

    International Nuclear Information System (INIS)

    Spore, J.W.; Lin, J.C.; Schnurr, N.M.; White, J.R.; Cappiello, M.C.

    1992-01-01

    Calculations of a large-break loss-of-coolant accident (LOCA) in a 15 x 15 generic four-loop Westinghouse nuclear power plant with both the TRAC-PF1/MOD1 and TRAC-PF1/MOD2 computer codes will be presented. The Transient Reactor Analysis Code (TRAC) has been developed by Los Alamos National Laboratory to provide advanced best-estimate simulations of real postulated transients in pressurized light-water reactors (LWRs) and for many related thermal-hydraulic facilities. The latest released version of TRAC is TRAC-PF1/MOD2. Significant improvements and enhancements over the MOD1 version were implemented in the MOD2 heat-transfer and constitutive models. One of the most significant improvements in the MOD2 code has been the implementation of the two-step numerics method in the three-dimensional components, which can significantly reduce run times for long, slow transients. A very important area of improvement has been in the reflood heat-transfer models. Developmental assessment results (i.e., code comparisons with experimental data) will be discussed for several separate-effects and integral test, including analysis of the Upper Plenum Test Facility (UPTF), the Cylindrical Core Test Facility (CCTF), and the Loss-of-Fluid Test Facility (LOFT). The assessment results provide information on the anticipated accuracy for the best-estimate models in the MOD2 computer code. The MOD1 to MOD2 comparison will provide an estimate for the effect of improved heat-transfer models on predicted peak cladding temperatures

  20. 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

  1. Assessment of selected TRAC and RELAP5 calculations for Oconee-1 pressurized thermal shock study

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Pu, J.; Saha, P.; Jo, J.

    1984-11-01

    Several Oconee-1 overcooling transients that were computed by LANL and INEL using the latest versions of TRAC-PF1 and RELAPS/MOD1.5 codes have been reviewed by BNL. Three of these transients were selected for detailed review as they either had the potential of challenging the integrity of the pressure vessel or highlighted the effect of code differences. These are: (1) Main Steam Line Break (MSLB); (2) All Turbine Bypass Valves Stuck Open; and (3) 2-Inch Small Break LOCA. Both codes were reasonably successful in modeling these transients. However, there were differences in the code results even though the specified scenarios were exactly the same for two transients (MSLB and Small Break LOCA). This report compares the code results and explains the possible reasons for these differences. Recommendations have been made regarding which result seems more reasonable for a specific transient

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Development of a hybrid safety system: Actuation of the secondary automatic depressurization system at an early stage

    International Nuclear Information System (INIS)

    Nishimoto, Masae; Umezawa, Shigemitsu; Okabe, Kazuharu; Matsuoka, Tsuyoshi

    1996-01-01

    A Hybrid Safety System, which is an optimum combination of active and passive safety systems, has been developed in order to improve the safety, reliability and economic features of the next generation of PWRs. The passive safety systems include Automatic primary Depressurization System (ADS), Secondary Automatic Depressurization System (SADS), advanced accumulators, gravity injection system and so on. In this study the authors have improved the actuation logic of the passive safety systems. The original logic in the previous study actuates ADS at an early stage of an event such as a Loss-of-Coolant Accident (LOCA), and this is followed by the actuation of SADS. In this study they divide SADS into two systems. The first, small SADS, uses small valves corresponding to the relief valves of the conventional PWR plants. The second, large SADS, corresponds to the original SADS using multiple valves of large capacity. With the new logic, the passive systems are actuated during a typical small LOCA. Small LOCA analyses using several break areas were performed for a 1,400 MWe PWR plant with a Hybrid Safety System. The results predict that core uncovery does not occur in the case of a relatively small break area and that core heat removal during a small LOCA is improved in comparison with the analyses for conventional PWR plants, where the secondary pressure remains higher during the event. The results also predict that this new logic make it possible to reduce the ADS valve size and the actuation pressure setpoint of the passive safety systems

  5. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications

  6. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

  7. A Theoretical Model for the Prediction of Siphon Breaking Phenomenon

    International Nuclear Information System (INIS)

    Bae, Youngmin; Kim, Young-In; Seo, Jae-Kwang; Kim, Keung Koo; Yoon, Juhyeon

    2014-01-01

    A siphon phenomenon or siphoning often refers to the movement of liquid from a higher elevation to a lower one through a tube in an inverted U shape (whose top is typically located above the liquid surface) under the action of gravity, and has been used in a variety of reallife applications such as a toilet bowl and a Greedy cup. However, liquid drainage due to siphoning sometimes needs to be prevented. For example, a siphon breaker, which is designed to limit the siphon effect by allowing the gas entrainment into a siphon line, is installed in order to maintain the pool water level above the reactor core when a loss of coolant accident (LOCA) occurs in an open-pool type research reactor. In this paper, we develop a theoretical model to predict the siphon breaking phenomenon. In this paper, a theoretical model to predict the siphon breaking phenomenon is developed. It is shown that the present model predicts well the fundamental features of the siphon breaking phenomenon and undershooting height

  8. A Theoretical Model for the Prediction of Siphon Breaking Phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Youngmin; Kim, Young-In; Seo, Jae-Kwang; Kim, Keung Koo; Yoon, Juhyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A siphon phenomenon or siphoning often refers to the movement of liquid from a higher elevation to a lower one through a tube in an inverted U shape (whose top is typically located above the liquid surface) under the action of gravity, and has been used in a variety of reallife applications such as a toilet bowl and a Greedy cup. However, liquid drainage due to siphoning sometimes needs to be prevented. For example, a siphon breaker, which is designed to limit the siphon effect by allowing the gas entrainment into a siphon line, is installed in order to maintain the pool water level above the reactor core when a loss of coolant accident (LOCA) occurs in an open-pool type research reactor. In this paper, we develop a theoretical model to predict the siphon breaking phenomenon. In this paper, a theoretical model to predict the siphon breaking phenomenon is developed. It is shown that the present model predicts well the fundamental features of the siphon breaking phenomenon and undershooting height.

  9. Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

    International Nuclear Information System (INIS)

    Khvostov, G.; Wiesenack, W.; Zimmermann, M.A.; Ledergerber, G.

    2011-01-01

    Highlights: → A model for the dynamics of axial gas redistribution in fuel rods during the LOCA is developed and coupled to the FALCON fuel behaviour code. → The first verification of the model is carried out using the data of the selected Halden LOCA tests. → According to calculation, the short rods used in the Halden tests show a small effect of the delayed gas redistribution during the clad ballooning. → The predicted effect is significant in the full length rods, eventually resulting in a considerable delay of the predicted moment of cladding rupture. → The predicted delay of cladding burst may be large enough to eventually affect the efficiency of the emergency core cooling system. - Abstract: A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code. The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from

  10. Phebus program main results and status for severe fuel damage studies

    International Nuclear Information System (INIS)

    Duco, J.; Reocreux, M.; Tattegrain, A.

    1986-06-01

    A large experimental in-pile program has been set up at the PHEBUS facility to investigate the actual behavior of .8 m active height, 25-rod PWR-type pressurized fresh fuel bundles under typical accident conditions. The program consists of four stages. Stage 1 was devoted to the adjustment of the operational procedure for stage 2. Stage 2 refers to the simulation of conservatively calculated L.B. LOCA 2 - peak transients. Stages 3/4 refer to four PWR severe accident scenarios retained for in-pile simulation at PHEBUS: a) a large break LOCA with injection failure; b) a small break LOCA associated with an injection failure; c) a prolonged total loss of the steam generator feedwater; and, d) a prolonged core uncovery a few days after reactor shutdown. The main PHEBUS stage 2 results are presented and finally interpreted

  11. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

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

  12. A role for small RNAs in DNA double-strand break repair

    DEFF Research Database (Denmark)

    Wei, W.; Ba, Z.; Wu, Y.

    2012-01-01

    Eukaryotes have evolved complex mechanisms to repair DNA double-strand breaks (DSBs) through coordinated actions of protein sensors, transducers, and effectors. Here we show that ∼21-nucleotide small RNAs are produced from the sequences in the vicinity of DSB sites in Arabidopsis and in human cells....... We refer to these as diRNAs for DSB-induced small RNAs. In Arabidopsis, the biogenesis of diRNAs requires the PI3 kinase ATR, RNA polymerase IV (Pol IV), and Dicer-like proteins. Mutations in these proteins as well as in Pol V cause significant reduction in DSB repair efficiency. In Arabidopsis, di...

  13. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  15. An analysis on boron dilution events during SBLOCA for the KNGR

    International Nuclear Information System (INIS)

    Kim, Young In; Hwang, Young Dong; Park, Jong Kuen; Chung, Young Jong; Sim, Suk Gu

    1999-02-01

    An analysis on boron dilution events during small break loss of coolant accident (LOCA) for Korea Next Generation Reactor (KNGR) was performed using Computational Fluid Dynamic (CFD) computer program FLUENT code. The maximum size of the water slug was determined based on the source of un borated water slug and the possible flow paths. Axisymmetric computational fluid dynamic analysis model is applied for conservative scoping analysis of un borated water slug mixing with recirculation water of the reactor system following small break LOCA assuming one Reactor Coolant Pump (RCP) restart. The computation grid was determined through the sensitivity study on the grid size, which calculates the most conservative results, and the preliminary calculation for boron mixing was performed using the grid. (Author). 17 refs., 3 tabs., 26 figs

  16. Possibilities of optimizing non-nuclear simulation of pressurized water reactor transients

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1985-01-01

    The GKSS-Forschungszentrum Geesthacht GmbH has instituted the concept of a scaled test facility (volume scale factor of 1/100) of a typical PWR of the 1 300 MWe class for the purpose of studying small breaks Loss-of-Coolant Accidents (LOCA) and transients. Having in mind the goal of an optimization of this concept has been choosen a station blackout with and without reactor shutdown and a small break LOCA in a primary loop piping to investigate the thermohydraulic behaviour of the test facility in comparison to the reactor plant. The computer code RELAP 5/MOD 1 has been utilized to compare the test facility behaviour with the reactor plant one. Recommendations are given for minimization of distortions between test facility and reactor plant. (orig./HP) [de

  17. Effects of the reactor coolant pumps following a small break in a Westinghouse PWR

    International Nuclear Information System (INIS)

    Koenig, J.E.

    1983-10-01

    Numerical simulations of the thermal-hydraulic events following a small cold-leg break in a Westinghouse pressurized water reactor were performed to address the pumps-on/off issue. The mode of pump operation was varied in each calculation to ascertain the optimum mode. It was found that pump operation was not critical for this break size and location because the fuel rods remained cool in all accidents analyzed. In terms of system mass, however, it was preferable to leave the pumps in operation

  18. Implementation of PWR steady state self-initialization feature into RELAP4/MOD6/U4/J3

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    1987-07-01

    A PWR steady state self-initialization feature has been implemented into the RELAP4/MOD6/U4/J3 code which is an improved version of RELAP4/MOD6 and can analyze not only large break but also small break LOCA in LWRs. This feature is originated from RELAP4/MOD7 which is the most updated released version of RELAP4 from INEL. Several FORTRAN subroutines in MOD7 related to this feature were transplanted into MOD6/U4/J3 with some improvements, which were the modification of method to take a balance of heat transfer between primary and secondary side at SG-U tubes, and to make it possible to nodalize secondary side of SG as multi-node. Advantages realized by implementation of this option are saving of time in initializaing a new model and an assurance of steady state and self consistency of input data in a small break LOCA analysis of a PWR. (author)

  19. Description of the small plastics fragments in marine sediments along the Alang-Sosiya ship-breaking yard, India

    Science.gov (United States)

    Srinivasa Reddy, M.; Basha, Shaik; Adimurthy, S.; Ramachandraiah, G.

    2006-07-01

    This study aimed to assess the accumulation of small plastic debris in the intertidal sediments of the world's largest ship-breaking yard at Alang-Sosiya, India. Small plastics fragments were collected by flotation and separated according to their basic polymer type under a microscope, and subsequently identified by FT-IR spectroscopy as polyurethane, nylon, polystyrene, polyester and glass wool. The morphology of these materials was also studied using a scanning electron microscope. Overall, there were on average 81 mg of small plastics fragments per kg of sediment. The described plastic fragments are believed to have resulted directly from the ship-breaking activities at the site.

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

    International Nuclear Information System (INIS)

    Shadday, M.A. Jr.

    1990-08-01

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

  1. Sensitivity of break-flow-partition on the containment pressure and temperature

    International Nuclear Information System (INIS)

    Kwon, Young Min; Song, Jin Ho; Lee, Sang Yong

    1994-01-01

    For the case of RCS blowdown into the vapor region of a containment at low pressure, the blowdown mixture will start to boil at the containment pressure and liquid will separate from the flow near the break location. The flashed steam is added to the containment atmosphere and liquid is falled to the sump. Analytically, the break flow can be divided into steam and liquid in a number of ways. Discussed in this study is three partition models and Instantaneous Mixing(IM) Model employed in different containment analysis computer codes. IM model is employed in the CONTRANS code developed by ABB-CE for containment thermodynamic analysis. The various partition models were applied to the double ended discharge leg slot break (DEDLS) LOCA which is containment design base accident (CDBA) for Ulchin 3 and 4 PSAR. It was shown that IM model is the most conservative for containment design pressure analysis. Results of the CONTRANS analyses are compared with those of UCN PSAR for which CONTEMPT-LT code was used

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. A study on the loss-of-coolant accidents associated with the lung-men nuclear power station

    International Nuclear Information System (INIS)

    Teng, J.T.; Hsu, C.T.; Wang, T.Q.; Chen, Y.H.; Wang, L.C.; Chung, N.M.; Yuann, R.Y.

    2001-01-01

    This study was intended to evaluate the behavior of the nuclear core of the Lung-Men Nuclear Power Station (LMNPS) under postulated LOCA conditions. The LMNPS construction is now in suspense by the Ministry of Economic Affairs, the Republic of China. The assumptions used in this study were in compliance with the requirements specified in 10CFR50.46 and Appendix K. The methodology used was primarily RELAP5YA, which was a modification to the RELAP5/MOD1 Cycle 18. In the paper, features of the thermo-fluids, neutronics, flow systems, trips, and breaks are discussed. Their assumptions and the resulting implications to the outcome of the analyses are emphasized. Also typical sequences of events, the reactor pressure vessel (RPV) pressure, temperature and water inventory transients, and the ultimate core heat-ups for a number of break sizes, ranging from small- to large-break LOCAs, are delineated. The results of this study indicated that for all cases studied, the peak cladding temperature (PCT) was 699.1 Celsius degrees (1290.4 F). This PCT was much lower than the upper temperature limit of 1204.4 Celsius degrees (2200 F) specified in the acceptance criterion of 10CFR50.46. It is to be noted that for all cases studied, the highest PCTs obtained occurred at 4 s after the initiation of the LOCAs. The reason for the occurrence of these PCTs was the internal pump trip, allowing the pump to coast down and the pump to reverse. The next PCTs, resulted from the LOCA, were observed to occur only for the LOCA cases with feedwater line breaks. It did not happen for the cases with steam-line breaks. (authors)

  4. Duality after supersymmetry breaking

    International Nuclear Information System (INIS)

    Shadmi, Yael; Cheng, Hsin-Chia

    1998-05-01

    Starting with two supersymmetric dual theories, we imagine adding a chiral perturbation that breaks supersymmetry dynamically. At low energy we then get two theories with soft supersymmetry-breaking terms that are generated dynamically. With a canonical Kaehler potential, some of the scalars of the ''magnetic'' theory typically have negative mass-squared, and the vector-like symmetry is broken. Since for large supersymmetry breaking the ''electric'' theory becomes ordinary QCD, the two theories are then incompatible. For small supersymmetry breaking, if duality still holds, the magnetic theory analysis implies specific patterns of chiral symmetry breaking in supersymmetric QCD with small soft masses

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. Improvement on models associated with LOCA and loss of RHR accidents during shutdown

    International Nuclear Information System (INIS)

    Chang, W. P.; Chung, Y. J.; Kim, W. S.; Kim, K. D.; Lee, S. J.; Jung, J. J.; Ha, G. S.; Son, Y. S.; Chung, B. D.; Han, D. H.; Lee, Y. J.; Hwang, T. S.; Lee, S. Y.; Park, C. Y.; Choi, H. R.; Lee, S. Y.; Choi, J. H.; Ban, C. H.; Bae, G. H.

    1997-07-01

    The characteristics of the best estimate codes available in Korea have been studied through literature surveys for the reliability on LOCA analyses and then, a feasibility study on reduction of capacities of existing safety systems in YGN 3/4 have been carried out using the codes. Since it has been expected to adopt DVI + 4 -Train HPSI in the next generation reactor, the core uncoveries under one DVI line break and 6 cold leg break, which is a requirement for advance d reactor by EPRI, in addition to LBLOCA for reduction effect of SIT capacity, have been analyzed. Finally, an effort on finding the way how the system could be simplified, has been made through the analysis of SIT injection characteristics. On the other hand, the best estimate methodology consisting of uncertainties of the code itself, bias, and application have been developed first and quantification of the uncertainty has been made the case of KORI unit 3 afterward. The prediction capabilities of the best estimate codes and major physical models on the accident under loss of RHR during shutdown have been assessed suing the large scale experimental data delivered from France and then, the assessed codes have been used to provide essential data required for description of operation procedures in YGN 3/4. (author). 64 refs., 45 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  8. 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

  9. 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

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

    International Nuclear Information System (INIS)

    1984-02-01

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

  11. Proceedings of the seminar on leak before break in reactor piping and vessels

    International Nuclear Information System (INIS)

    Faidy, C.; Gilles, P.

    1997-04-01

    The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  12. Proceedings of the seminar on leak before break in reactor piping and vessels

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [ed.] [Electricite de France, Villeurbanne (France); Gilles, P. [ed.] [Framatome, Paris (France)

    1997-04-01

    The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  13. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  16. Strategies of modeling the cognitive tasks of human operators for accident scenarios in nuclear power plant control rooms

    International Nuclear Information System (INIS)

    Cheon, Se Woo; Sur, Sang Moon; Lee, Yong Hee; Lee, Jeong Wun

    1993-01-01

    This paper presents the development strategies of cognitive task network modeling for accident scenarios in nuclear power plant control rooms. Task network modeling is used to provide useful predictions of operator's performance times and error rates, based upon plant procedures and/or control room changes. Two accident scenarios, small-break loss of coolant accident (LOCA) and steam generator tube rupture (SGTR), are selected for task simulation. To obtain the input data for the model, task elements are extracted by the task analysis of emergency operating procedures. The input data include task performance time, communication ink, panel location, component operating mode, and data for performance shaping factors (PSFs). Operator's verbs are categorized according to the elements of cognitive behavior. The simulation of the task network for the small-break LOCA scenario is presented in this paper. (Author)

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

    International Nuclear Information System (INIS)

    Jang, Chansu; Um, Kilsup; Choi, Jaedon

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. The common project for completion of Bubbler Condenser Qualification (Bohunice, Mochovce, Dukovany and Paks NPPs)

    International Nuclear Information System (INIS)

    Jaroslav, H.; Pavol, B.

    2003-01-01

    Described is the common project for completion of bubbler condenser qualification for nuclear power plants in Bohunice, Mochovice, Dukovany and Paks. Functionality of the bubbler condenser was elaborated during the simulation of the main steam line brake, medium break and small break LOCA. On this basis the appropriate operation of bubbler condenser containment under accident conditions can be positively confirmed

  20. TRAC-PF1/MOD2 status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Steinke, R.G.; Nelson, R.A.; Cappiello, M.W.; Jenks, R.

    1989-01-01

    The development of the TRAC-PF1/MOD1 code was completed in July 1988 with the release of Version 14.4. A TRAC-PF1/MOD2 code development plan addresses code deficiencies identified in the MOD1 code in order to provide an accurate and defensible tool that can be used to simulate large-break loss-of-coolant accidents (LOCAs), small-break LOCAs, and operational transients. The MOD2 code development plan is an international cooperative effort that includes contributions from Los Alamos National Laboratory, Idaho National Engineering Laboratory (INEL), Japanese Atomic Energy Research Institute (JAERI), Cray Research, Central Electricity Generating Board (CEGB), and United Kingdom Atomic Energy Authority (UKAEA)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  3. Simulation of the SPE-4 small-break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Cebull, P.; Hassan, Y.A.

    1993-01-01

    A small-break loss of coolant accident (SBLOCA) conducted at the PMK-2 integral test facility was analyzed using RELAP5/MOD3. 1. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). The VVER design differs from pressurized water reactors (PWRS) of western origin, primarily in its use of horizontal steam generators, hot- and cold-leg loop seals, and safety injection tanks. Because of these differences, it will exhibit somewhat different transient behavior than most PWRS. The PMK-2 test facility, located at the KFKI Atomic Energy Research Institute (AEKI), is a scale model of the Paks nuclear power plant in Hungary with scaling factors of 1:2070 in power and volume and 1:1 in elevation. Primarily used to study SBLOCAs and natural circulation behavior of VVER reactors, it has been used in three previous SPEs

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  6. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  7. TRAC-PF1/MOD1 computer code

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1984-01-01

    The TRAC-P1 program was designed primarily for the analysis of large-break loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs). Because of its versatility, however, it can be applied directly to many analyses ranging from blowdowns in simple pipes to integral LOCA tests in multiloop facilities. A refined version, called TRAC-P1A, was released to the National Energy Software Center (NESC) in March 1979. Although it still treats the same class of problems, TRAC-P1A is more efficient than TRAC-P1 and incorporates improved hydrodynamic and heat-transfer models. It also is easier to implement on various computers. TRAC-PD2 contains improved reflood and heat-transfer models and improvements in the numerical solution methods. Although a large LOCA code, it has been applied successfully to small-break problems and to the Three Mile Island incident. Distinguishing characteristics of the TRAC-PF1/MOD1 are summarized

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

    International Nuclear Information System (INIS)

    Hazen Hezhi Fan; Zoran Bilanovic

    2005-01-01

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

  9. Pipe rupture test results; 6 in. pipe whip test under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi; Yano, Toshikazu; Ueda, Shuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kato, Rokuro; Miyazono, Shohachiro

    1983-02-01

    A series of pipe rupture tests has been performed in JAERI to demonstrate the safety of the primary coolant circuits in the event of pipe rupture, in nuclear power plants. The present report summarizes the results of 6 in. pipe whip tests (RUN 5605, 5606), under BWR LOCA conditions (285 0 C, 6.8 MPa), which were performed in August, 1981. The test pipe is made of Type 304 stainless steel and its outer diameter is 6 in. and its thickness is 11.1 mm. The restraints are made of Type 304 stainless steel and its diameter is 16.0 mm. Two restraints were set on the restraint support with clearance of 100 mm. Overhang length was varied as the parameter in these tests and was 300 mm or 700 mm. The following results are obtained. (1) The deformations of a pipe and restraints are limited effectively by shorter overhang length of 300. However, they become larger when the overhang length is 700 mm, and the pipe deforms especially at the setting point of restraints. (2) Velocity at the free end of pipe becomes about 30 m/sec just after the break. However, velocity at the setting point of restraint becomes about only 4 m/sec just after the break. (3) It seems from the comparison between the 4 in. tests and 6 in. tests that the maximum restraint force of 6 in. tests is about two times as large as that of 4 in. tests. (author)

  10. Scaling criteria and an assessment of Semiscale Mod-3 scaling for small-break loss-of-coolant transients

    International Nuclear Information System (INIS)

    Larson, T.K.; Anderson, J.L.; Shimeck, D.J.

    1982-01-01

    Various methods of scaling fluid thermal-hydraulic test facilities and their relative merits and disadvantages are examined in light of nuclear reactor safety considerations. Particular emphasis is placed on examination of the scaling of the Semiscale Mod-3 system and determination of thermal-hydraulic phenomena thought to be important during a small break loss-of-coolant accident in a pressurized water nuclear reactor. The influence of geometric and dynamic scaling concerns in the Mod-3 system on small break behavior are addressed from an engineering viewpoint and corrective measures contemplated or required to make results from Semiscale tests more meaningful relative to expected PWR response are discussed

  11. THYDE-P, PWR LOCA Thermohydraulic Transient Analysis

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2001-01-01

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

  12. Development of risk-informed system design methodology for future nuclear power plants

    International Nuclear Information System (INIS)

    Yang, J. H.; Kim, M. R.; Park, S. J.; Lim, H. K.; Ji, S. K.; Choi, C. J.

    2002-01-01

    The purpose of this analysis is to develop the risk assessment evaluation process that can reduce the conservatism involved in the LOCA quantification. The frequency estimation for LOCA was performed according to NUREG/CR-5750. The raw data for LOCA events described in NUREG/CR-5750 was applied to this project. Lots of thermal hydraulic analyses for various break sizes were performed to find the boundary conditions that can effect the success criteria of event mitigation. The MARS 2.1 code, best-estimated computer code, was used in this analysis. The analysis result shows that conservatism in the LOCA quantification can be reduced when the detailed LOCA breakdown supported thermal hydraulic analysis is performed in the PSA model. The CDF for new re-classified LOCA events was reduced about 50% of current model's. Concurrent with the LOCA re-classification, the operator's available time for the feed and bleed operation using Safety Depressurization System (SDS) valves during small LOCA and its contribution to CDF were considered. Its results did not have an effect of CDF reduction, but it is believed that the iterative approach and findings are very useful

  13. 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

  14. 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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  16. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  19. 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

  20. Lesson learned from the application to LOBI tests of CATHARE and RELAP5 codes

    International Nuclear Information System (INIS)

    Ambrosini, W.; D'Auria, F.; Galassi, G.M.

    1992-01-01

    The Dipt. di Costruzioni Meccaniche e Nucleari has participated to the LOBI project since its very beginning, contributing to almost all the international activities in this field, such as task group meetings, International Standards Problems, Seminars, etc. System codes like RELAP4/MOD6, RELAP5/MOD1, RELAP5/MOD1-EUR, RELAP5/MOD2, CATHARE 1 and CATHARE 2 were applied to the design and post test evaluation of a wide series of both LOBI/MOD1 and LOBI/MOD2 experiments, including Large Break LOCAs, Small and Intermediate Break LOCAs, long lasting transients and characterization tests. The LOBI data base demonstrated its usefulness in assessing capabilities and limitations of these codes and in qualifying a code use strategy. (author)

  1. 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

  2. Analysis of three ex-vessel loss-of-coolant accidents in the first wall cooling system of NET/ITER

    International Nuclear Information System (INIS)

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

    1993-01-01

    An ex-vessel LOCA may be caused by a rupture of a cooling pipe located outside the vacuum vessel. No plasma shutdown and no other counteractions have been assumed in order to study the worst case conditions of the accidents. The next three ex-vessel LOCAs in the primary cooling system of the first wall have been analysed: 1. a large break ex-vessel LOCA caused by a rupture of the cold leg (inner diameter 0.314 m) of the main circuit; 2. an intermediate break ex-vessel LOCA caused by a rupture of a sector inlet feeder (inner diameter 0.158 m); 3. an intermediate break ex-vessel LOCA caused by a rupture of the surge line (inner diameter 0.180 m) of the pressurizer. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the first two scenarios, melting in the first wall starts about 90 s after break initiation. In the third scenario, melting in the first wall start about 323 s after break initiation. Special emphasis has been paid to the characteristics of the break flows, the transient thermal-hydraulic behaviour of the cooling system, and the temperature development in the first wall. (orig.)

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

    International Nuclear Information System (INIS)

    Maertens, D.; Spykman, G.

    2005-01-01

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

  4. Suggestion for a homogenizer installation in LOFT small break two-phase measurement

    International Nuclear Information System (INIS)

    Rieger, G.

    1981-07-01

    The purpose of this task, which was performed as an Austrian inkind contribution for the INEL research program is a) the evaluation of literature concerning homogenizers to improve two phase flow measurements for the LOFT small break test series, b) design of a homogenizer and c) recommandation of the location of a homogenizer in the LOFT piping system. To optimize the location of the homogenizer LTSF-tests should be performed according to the suggestions in this paper. (author)

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

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  6. Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

    International Nuclear Information System (INIS)

    Kolev, N.I.; Sybotinov, L.S.

    1984-01-01

    The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

  7. Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

    This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

  8. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  9. 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.

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

    International Nuclear Information System (INIS)

    Lee, Joosuk; Woo, Swengwoong

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  12. TRAC-PF1/MOD1 assessment at Los Alamos

    International Nuclear Information System (INIS)

    Knight, T.D.

    1984-01-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in pressurized water reactors (PWRs). Over the past several years, four distinct versions of the code have been released; each new version introduced improvements to the existing models and numerics and added new models to extend the applications of the code. The first goal of the code was to analyze large-break loss-of-coolant accidents (LOCAs), and the TRAC-P1A and TRAC-PD2 codes primarily addressed the large-break LOCA. (The TRAC-PD2/MOD1 code is essentially the same as the TRAC-PD2 code but it also includes a released set of error corrections.) The TRAC-PF1 code contained major changes to the models and trips and to the numerical methods. These modifications enhanced the computational speed of the code and improved the application to small-break LOCAs. The TRAC-PF1/MOD1 code, the latest released version, added improved steam-generator modeling, a turbine component, and a control system together with modified constitutive relations to model the balance of plant on the secondary side and to extend the applications to non-LOCA transients. The TRAC-PF1/MOD1 code also contains reasonably general reactor-kinetics modeling to facilitate the simulation of transients with delayed scram or without scram. 13 references, 24 figures

  13. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  14. Transient analysis of the IRIS reactor

    International Nuclear Information System (INIS)

    Bajs, T.; Oriani, L.; Ricotti, M.E.; Barroso, A.C.

    2002-01-01

    An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a modular, integral, light water cooled, small to medium power reactor, the International Reactor Innovative and Secure (IRIS). IRIS features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). Given the large number of organizations involved in the IRIS design, the RELAP5/MOD 3.3 code has been selected as the main system code. A nodalization of the reference IRIS design has been developed with a basic set of protective functions and controls. Engineered Safety Features of the concept are being also implemented, and in particular the Emergency Heat Removal System that is used for safety grade decay heat removal and in the small break LOCA response of IRIS (Large break LOCAs are eliminated in IRIS by the adoption of the Integral layout) This paper discusses developed model and transient behavior of the system for representative transient sequences.(author)

  15. 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

  16. FLECHT SEASET program. Final report

    International Nuclear Information System (INIS)

    Hochreiter, L.E.

    1985-11-01

    This report presents the highlights and main findings of the USNRC, EPRI, and Westinghouse cooperative FLECHT SEASET program. The report indicates areas in which the results of the program can contribute to revising the current licensing requirements for Loss of Coolant (LOCA) safety analysis for PWRs. Also identified are several technical areas in which the new FLECHT SEASET data and analysis can lead to improved safety analysis modeling, and thereby to predicted PWR response for postulated accident scenarios. Significant progress has been made in the modeling areas of nonequilibrium dispersed two-phase flow during reflood. Improved models and understanding of this rod bundle cooling regime are summarized in this report. Another important result of the FLECHT SEASET program arises from the natural circulation test series, which investigated single-phase, two-phase, and reflux condensation cooling modes of a scaled PWR under small-break LOCA conditions. The tests and subsequent analysis constitute one of few complete sets of data for these cooling modes in which full-height, multitube steam generators with sufficient instrumentation were used to examine primary-to-secondary heat transfer in the generators. It is believed that the natural circulation test data will be extremely useful to benchmark the improved post-TMI small-break LOCA computer codes. 170 figs., 13 tabs

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  18. LOCA scenario tests of irradiated fuel rod specimens

    International Nuclear Information System (INIS)

    Scott, Harold

    2004-01-01

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

  19. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  20. Study of Dam-break Due to Overtopping of Four Small Dams in the Czech Republic

    Directory of Open Access Journals (Sweden)

    Zakaraya Alhasan

    2015-01-01

    Full Text Available Dam-break due to overtopping is one of the most common types of embankment dam failures. During the floods in August 2002 in the Czech Republic, several small dams collapsed due to overtopping. In this paper, an analysis of the dam break process at the Luh, Velký Bělčický, Melín, and Metelský dams breached during the 2002 flood is presented. Comprehensive identification and analysis of the dam shape, properties of dam material and failure scenarios were carried out after the flood event to assemble data for the calibration of a numerical dam break model. A simple one-dimensional mathematical model was proposed for use in dam breach simulation, and a computer code was compiled. The model was calibrated using the field data mentioned above. Comparison of the erodibility parameters gained from the model showed reasonable agreement with the results of other authors.

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

    International Nuclear Information System (INIS)

    Jonsson, T.

    1978-10-01

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

  2. MELCOR based severe accident simulation for WWER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Vegh, E.; Buerger, L.; Gacs, A.; Gyenes, F.G.; Hozer, Z.; Makovi, P.

    1997-01-01

    SUBA is a MELCOR based severe accident simulator, installed this summer at the Hungarian Nuclear Safety Directorate. In this simulator the thermohydraulics, chemical reactions and material transport in the primary and secondary systems are calculated by the MELCOR code, but the containment, except the cavity, is modelled by the HERMET code, developed in our Institute. The instrumentation and control, the safety systems and the plant logic, are calculated by our models. This paper describes the main features of the used models and presents three different test transients. The presented transients are as follows: a small break LOCA, a cold leg large break LOCA, and the station blackout, without Diesel generators. In each treated transients the most important parameters are presented as time functions and the most significant events are analysed. (author)

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

    International Nuclear Information System (INIS)

    Osmachkin, V.S.

    1999-01-01

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

  4. SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

    2012-10-15

    APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

  5. 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

  6. RELAP5/MOD2 post-test calculation of the OECD LOFT experiment LP-SB-2

    International Nuclear Information System (INIS)

    Perez, J.; Mendizabal, R.

    1992-04-01

    This document presents the analysis of the OECD LOFT LP-SB-2 Experiment performed by the Consejo de Seguridad Nuclear of Spain working group making use of RELAP5/MOD2 in the frame of the Spanish LOFT Project. LB-SB-2 experiment studies the effect of a delayed pump trip in a small break LOCA scenario with a 3-inch equivalent diameter break in the hot leg of a commercial PWR

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  8. PWR small-break analysis using a PDP-11/AD10 computer system

    International Nuclear Information System (INIS)

    Venhuizen, J.R.; Hyer, F.K.

    1983-01-01

    A simulation of a pressurized water test reactor was developed to predict the dynamic response of the primary coolant system to gradual voiding caused by an anticipated transient or a small break. Comparison of the simulation results with data from the LOFT test reactor at the Idaho National Engineering Laboratory was performed to verify the models. The simulation, designed to operate on a PDP-11/55 minicomputer and Applied Dynamic AD10 synchronous digital computer, was used interactively to do scoping analysis prior to running the transient at the test reactor

  9. Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Isozaki, Toshikuni; Yano, Toshikazu; Miyazaki, Noriyuki; Kato, Rokuro; Kurihara, Ryoichi; Ueda, Shuzo; Miyazono, Shohachiro

    1982-09-01

    Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

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

    International Nuclear Information System (INIS)

    Erbacher, F.J.

    1982-01-01

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

  11. Review of boiling water reactor small break loss of coolant accidents

    International Nuclear Information System (INIS)

    Gururaj, P.M.; Dua, S.S.; Rao, A.S.

    1981-01-01

    This paper presents a review of the analytical and the experimental work performed by the General Electric Company to determine the performance of boiling water reactors (BWR) following postulated small break accidents (SBA). This review paper addresses the following issues: (1) the response of the BWR following small loss of inventory events; (2) methods of analysis and their justification; (3) necessity, if any, of operator action and the length of time available in which such action can be performed; and (4) operator interface following the SBA event. The results from these SBA studies for different BWR product lines show that even with the multiple system failures assumed, the BWR can successfully withstand an SBA. For a typical BWR/6, it takes the failure of 13 water delivery pumps to cause any significant core heatup. The only operator actions determined to be necessary are simple ones and ample time is available to the operator to perform these actions, if needed

  12. Stratification studies in components of nuclear power plants

    International Nuclear Information System (INIS)

    Randorf, J.A.

    1997-01-01

    The applicability of two stratification criteria during loss-of-coolant (LOCA) conditions was studied. The first criteria was developed for addressing cold water injection-induced stratification. The second criteria applied to downcomer/cold leg junction stratification. Both criteria provided predictions consistent with measured conditions during small break loss-of-coolant tests

  13. Review of RIA and LOCA criteria for WWER fuel

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

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

  14. Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

    International Nuclear Information System (INIS)

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

    1998-01-01

    Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. the applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous's empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing

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

    International Nuclear Information System (INIS)

    Reocreux, M.

    1980-05-01

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

  16. Small extra dimensions from the interplay of gauge and supersymmetry breaking

    International Nuclear Information System (INIS)

    Buchmueller, W.; Catena, R.; Schmidt-Hoberg, K.

    2008-03-01

    Higher-dimensional theories provide a promising framework for unified extensions of the supersymmetric standard model. Compactifications to four dimensions often lead to U(1) symmetries beyond the standard model gauge group, whose breaking scale is classically undetermined. Without supersymmetry breaking, this is also the case for the size of the compact dimensions. Fayet-Iliopoulos terms generically fix the scale M of gauge symmetry breaking. The interplay with supersymmetry breaking can then stabilize the compact dimensions at a size 1/M, much smaller than the inverse supersymmetry breaking scale 1/μ. We illustrate this mechanism with an SO(10) model in six dimensions, compactified on an orbifold. (orig.)

  17. 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

  18. A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

    International Nuclear Information System (INIS)

    Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

    2002-01-01

    A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-12-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  1. Thermal fluid mixing behavior during medium break LOCA in evaluation of pressurized thermal shock

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jae Won; Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    1998-12-31

    Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze the thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of thermal stratification is investigated using Theofanous`s empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing. 6 refs., 8 figs. (Author)

  2. 75 FR 32509 - Notice Applications and Amendments to Facility Operating Licenses Involving Proposed No...

    Science.gov (United States)

    2010-06-08

    ... S-RELAP5 for the Final Safety Analysis Report (FSAR) Chapter 15 realistic large- break LOCA in the... a request or petition for hearing (even in instances in which the participant, or its counsel or... Commission (NRC)-approved topical report (TR) EMF-2103(P)(A), Revision 0, ``Realistic Large-Break LOCA [Loss...

  3. 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

  4. An Analysis of Effect of Break-up Timing on the Necessity of a Feed-and-Bleed Operation in the case of TLOFW with Local

    International Nuclear Information System (INIS)

    Kim, Bo Gyung; Kim, Sang Ho; Kang, Hyun Gook; Yoon, Ho Joon

    2014-01-01

    A Feed-and-bleed (F and B) operation is a process to cool the reactor by the primary side directly. If adequate residual heat removal through the secondary side is not available, the heat can be removed from the RCS by F and B operation. A total loss of feedwater (TLOFW) accident is used to represent an accident involving the failure of cooling by the secondary cooling system. Even if the secondary cooling system fails, the RCS can be cooled by F and B transients when a loss of coolant accident (LOCA) with a TLOFW accident occurs. During an F and B transient, the RCS has a residual heat removal mechanism. If the break size is large, an F and B transient continuously occurs if the SIS is available. If the break size is small to sufficiently decrease the RCS pressure, the SIS cannot inject the coolant, causing the F and B transient to terminate. After the termination of the F and B transient, the residual heat cannot be removed, and the necessity of an F and B operation increases. The operators may hesitate to initiate F and B operation if a clear cue is not provided, since its initiation implies the radioactive coolant releases into the containment. Therefore, the necessity of F and B operation is needed to be identified. The factors affected the necessity of F and B operation are the availability of the safety injection system and safety depressurization system, water inventory in the primary and secondary cooling systems, break size in a loss-of-coolant accident, and time of accident occurrence. The necessity of F and B operation can be changed according to different timing of break-up despite same break size. Moreover, different timing of break-up makes the operators more complicated. To identify effect of timing of break-up, a thermohydraulic analysis was performed using the MARS code. This study is expected to provide a useful guideline to identify the necessity of an F and B operation under combined accident

  5. 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.

  6. 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

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

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  8. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Zhang-cheng; Xie, Heng

    2014-01-01

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  9. A probabilistic SSYST-3 analysis for a PWR-core during a large break LOCA

    International Nuclear Information System (INIS)

    Schubert, J.D.; Gulden, W.; Jacobs, G.; Meyder, R.; Sengpiel, W.

    1985-05-01

    This report demonstrates the SSYST-3 analysis and application for a German PWR of 1300 MW. The report is concerned with the probabilistic analysis of a PWR core during a loss-of-coolant accident due to a large break. With the probabilistic analysis, the distribution functions of the maximum temperatures and cladding elongations occuring in the core can be calculated. Parameters like rod power, the thermohydraulic boundary conditions, stored energy in the fuel rods and the heat transfer coefficient were found to be the most important. The expected value of core damage was determined to be 2.9% on the base of response surfaces for cladding temperature and strain deduced from SSYST-3 single rod results. (orig./HP) [de

  10. Large-break LOCA studies. Computational analysis of clad ballooning and thermohydraulics in a PWR

    International Nuclear Information System (INIS)

    Ammirabile, L.; Walker, S.

    2002-01-01

    A new multi-pin model of the re-flood phase of a large break loss of coolant accident has been created through the dynamic coupling between the thermal-hydraulic code RELAP5 and multiple instances of the single-pin thermal-mechanics code MABEL. After a brief description of the codes and their linkage, a series of tests to assess the capabilities of the linked codes is described, and their results analysed. It is shown that the current coupled multi-pin code is a stable and reliable tool for ballooning transient analysis. A complete validation process with the simulation of the MT-3 test in the NRU reactor at Chalk River is in progress.(author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  12. CATHARE2 calculation of SPE-3 test small break loca on PMK facility

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, E.; Radet, J. [Institut de Protection et de Surete Nucleaire, Cadarache (France)

    1995-09-01

    Bind and post test calculations with CATHARE2 have been performed concerning the SPE-4 exercise organized under the auspices of IAEA on the hungarian PMK-2 facility, a one loop scaled model of VVER 440/213 Nuclear Power Plant. The SPE-4 test is a cold leg SBLOCA associated to a {open_quotes}bleed and feed{close_quotes} procedure applied in the secondary circuit. The present paper is devoted to the analysis of the post test calculation. For the first part of the transient (until the end of the SIT activations), the primary and secondary pressures are rather well predicted, leading to a good agreement with the experimental trips, as scram, flow coast down, SIT beginning and end of activation. Nevertheless, some discrepancy with the experiment may be due to an over prediction of the thermal exchanges from the primary to the secondary circuits. For the second part of the transient, the predicted primary circuit repressurization is shifted after the SITs are off, while in the experiment this event immediately follows the end of SIT activation. The delay in the calculation leads to underpredict primary and secondary pressures, thus anticipating the timing of events, such as LPIS and emergency feedwater activation.

  13. 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

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

    International Nuclear Information System (INIS)

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

    1980-09-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  16. Development of the MARS input model for Ulchin 1/2 transient analyzer

    International Nuclear Information System (INIS)

    Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.

    2003-03-01

    KAERI has been developing the NSSS transient analyzer based on best-estimate codes for Ulchin 1/2 plants. The MARS and RETRAN code are used as the best-estimate codes for the NSSS transient analyzer. Among the two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the input model requirements and the calculation note for the Ulchin 1/2 MARS input data generation (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 1/2

  17. Development of the MARS input model for Ulchin 3/4 transient analyzer

    International Nuclear Information System (INIS)

    Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Hwang, M. G.

    2003-12-01

    KAERI has been developing the NSSS transient analyzer based on best-estimate codes.The MARS and RETRAN code are adopted as the best-estimate codes for the NSSS transient analyzer. Among these two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the MARS input model requirements and the calculation note for the MARS input data generation (see the Appendix) for Ulchin 3/4 plant analyzer. In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 3/4

  18. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  20. Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

    2012-01-01

    Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

  1. 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)

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-12-01

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

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

    International Nuclear Information System (INIS)

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

    1994-09-01

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

  5. The analysis of 14.8 percent cold leg break without the application of hydroaccumulators in the PMK-NHV test facility

    International Nuclear Information System (INIS)

    Szabados, L.; Ezsoel, Gy.; Perneczky, L.

    1990-12-01

    A series of reactor safety tests have been performed in the experimental reactor simulation facility PMK-NHV of the Paks Nuclear Power Plant, Hungary, with and without the use of hydroaccumulator (SIT) operation. 14.8 percent cold leg break simulation experiments are reported without SITs in action, and the measurement results were analyzed using the RELAP5/mod2 computer code. The description of the experiment is followed by the parameter variations and their analysis, together with an interpretation of the measurement results. The lessons from the LOCA simulation tests are evaluated. (R.P.) 10 refs.; 48 figs.; 3 tabs

  6. Uncertainty analysis of minimum vessel liquid inventory during a small-break LOCA in a B ampersand W Plant: An application of the CSAU methodology using the RELAP5/MOD3 computer code

    International Nuclear Information System (INIS)

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

    1992-12-01

    The Nuclear Regulatory Commission (NRC) revised the emergency core cooling system licensing rule to allow the use of best estimate computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability, and Uncertainty (CSAU) to evaluate best estimate code uncertainties. The objective of this work was to adapt and demonstrate the CSAU methodology for a small-break loss-of-coolant accident (SBLOCA) in a Pressurized Water Reactor of Babcock ampersand Wilcox Company lowered loop design using RELAP5/MOD3 as the simulation tool. The CSAU methodology was successfully demonstrated for the new set of variants defined in this project (scenario, plant design, code). However, the robustness of the reactor design to this SBLOCA scenario limits the applicability of the specific results to other plants or scenarios. Several aspects of the code were not exercised because the conditions of the transient never reached enough severity. The plant operator proved to be a determining factor in the course of the transient scenario, and steps were taken to include the operator in the model, simulation, and analyses

  7. Experimental Studies for the VVER-440/213 Bubble Condenser System for Kola NPP at the Integral Test Facility BC V-213

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, O.I.; Melikhov, V.I.; Davydov, M.V.; Wolff, H.; Arndt, S.

    2012-01-01

    In the frame of Tacis Project R2.01/99, which was running from 2003 to 2005, the bubble condenser system of Kola NPP (unit 3) was qualified at the integral test facility BC V-213. Three LB LOCA tests, two MSLB tests, and one SB LOCA test were performed. The appropriate test scenarios for BC V-213 test facility, modeling accidents in the Kola NPP unit 3, were determined with pretest calculations. Analysis of test results has shown that calculated initial conditions and test scenarios were properly reproduced in the tests. The detailed posttest analysis of the tests performed at BC V-213 test facility was aimed to validate the COCOSYS code for the calculation of thermohydraulic processes in the hermetic compartments and bubble condenser. After that the validated COCOSYS code was applied to NPP calculations for Kola NPP (unit 3). Results of Tacis R2.01/99 Project confirmed the bubble condenser functionality during large and small break LOCAs and MSLB accidents. Maximum loads were reached in the LB LOCA case. No condensation oscillations were observed.

  8. Report of the Bulletins and Orders Task Force. Volume II. Appendices

    International Nuclear Information System (INIS)

    1980-01-01

    Appendices include: Office of Inspection and Enforcement bulletins; NRR status report on feedwater transients in BWR plants; orders on Babcock and Wilcox Company plants; letters lifting orders; letters issuing auxiliary feedwater system requirements; letter to licensees of all operating reactors, dated October 30, 1979 concerning short-term lessons learned requirements; and letters approving guidelines for preparation of small-break LOCA operating procedures

  9. Data report of ROSA/LSTF experiment SB-CL-32. 1% cold leg break LOCA with SG depressurization and no gas inflow

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2014-11-01

    An experiment SB-CL-32 was conducted on May 28, 1996 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-CL-32 simulated a 1% cold leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and no inflow of non-condensable gas from accumulator (ACC) tanks of emergency core cooling system. Secondary-side depressurization of both steam generators (SGs) as an accident management (AM) action to achieve the depressurization rate of 200 K/h in the primary system was initiated 10 min after the break. After the initiation of AM action, auxiliary feedwater injection into the SG secondary-side was started with some delay. After the onset of AM action, the primary pressure decreased following the SG secondary-side pressure. Core uncovery by core boil-off started with liquid level drop in crossover leg downflow-side. The core liquid level recovered rapidly after first loop seal clearing (LSC). The surface temperature of simulated fuel rod then increased up to 669 K. Core uncovery by core boil-off took place before second LSC induced by steam condensation on ACC coolant injected into cold legs following the primary depressurization. The core liquid level recovered rapidly after the second LSC. The observed maximum fuel rod surface temperature was 772 K. The experiment was terminated when the continuous core cooling was confirmed because of the coolant injection by low pressure injection system after the isolation of ACC system. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal-hydraulic phenomena. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment SB-CL-32. (author)

  10. Experimental study on size effect of siphon-breaking hole in the real-scaled reactor pool

    International Nuclear Information System (INIS)

    Kang, Soon Ho; Ahn, Ho Seon; Kim, Ji Min; Kim, Moo Hwan; Lee, Kwon Yeong; Seo, Kyoung Woo; Chi, Dae Young

    2012-01-01

    A rupture in the primary piping of a cooling system with a heat source or in a research reactor could lead to a loss-of-coolant accident (LOCA). However, if the water level of the reactor pool could be sustained and a reactor scram follows, the heat source could be cooled by natural convection, and significant accidents could be avoided. When a piping-system rupture accident occurs, the coolant starts to siphon out of the reactor pool until the pressure head between the inlet and outlet is removed or the siphon flow is interrupted. Therefore, a siphon-breaker mechanism can be adopted as a passive safety device to maintain the reactor water level. The gas entrainment is used to block the continuous loss of coolant by interrupting the siphon flow. Siphon breaking is complicated due to the transient, turbulent, two-phase flow mode, so suitable models or correlations that describe this phenomenon do not exist, and no general analysis been developed. Previous researchers have conducted experiments and numerical simulations to design a siphon breaker to meet their needs. Previous research on siphon breaking has not been conducted systemically, and no literature exists, even though the topic is greatly concerned with hydraulic safety. In this study, siphon-breaking holes were used as siphon breakers, and their performance was evaluated by the residual water quantity. Flow visualization was conducted to interpret the siphon-breaking phenomenon

  11. Progress in realistic LOCA analysis

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  12. 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

  13. Pipe rupture test results: 4-inch pipe whip tests under PWR LOCA conditions

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki; Ueda, Shuzo; Isozaki, Toshikuni; Kato, Rokuro; Kurihara, Ryoichi; Yano, Toshikazu; Miyazono, Shohachiro

    1982-09-01

    This report summarizes the results of 4-inch pipe whip tests (RUN No. 5506, 5507, 5508 and 5604) under the PWR LOCA conditions. The dynamic behaviors of the test pipe and restraints were studied in the tests. In the tests, the gap between the test pipe and the restraints was kept at the constant value of 8.85 mm and the overhang length was varied from 250 mm to 650 mm. The dynamic behaviors of the test pipe and the restraint were made clear by the outputs of strain gages and the measurements of residual deformations. The data of water hammer in subcooled water were also obtained by the pressure transducers mounted on the test pipe. The main conclusions obtained from the tests are as follows. (1) The whipping of pipe can be prevented more effectively as the overhang length becomes shorter. (2) The load acting on the restraint-support structure becomes larger as the overhang length becomes shorter. (3) The restraint farther from the break location does not limit the pipe movement except for the first impact when the overhang length is long. (4) The ultimate moment M sub(u) of the pipe at the restraint location can be used to predict the plastic collapse of the whipping pipe. (5) The restraints slide along the pipe axis and are subjected to bending moment, when the overhang length is long. (author)

  14. 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.

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

    International Nuclear Information System (INIS)

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

    1980-05-01

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

  16. Reactor safety issues resolved by the 2D/3D Program

    International Nuclear Information System (INIS)

    Damerell, P.S.; Simons, J.W.

    1993-07-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

  17. Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

    International Nuclear Information System (INIS)

    Moses Lemuel Raj, G.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    1998-12-01

    Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

  18. Establishment of the operating procedure to prevent boron precipitation during Post-LOCA long term cooling for Korean Westinghouse 3-loop NPPs

    International Nuclear Information System (INIS)

    Choi, Han Rim; Kwon, Tae Soon; Ban, Chang Hwan; Jeong, Jae Hoon; Lee, Young Jin.

    1996-11-01

    During post-LOCA LTC the increase of the excess reactivity for the extended fuel cycle should require increasing the RWST boron concentration in order to ensure core subcritical state. To quantify the concentration increment, the calculation methods was developed for the post-LOCA RCS/Sump mixed mean boron concentration, which applied for Kori 3 and 4 and Ulchin 1 and 2 of the Westinghouse 3-loop nuclear power plants in Korean. From the calculation results, the minimum boric acid concentrations increased of the RWST and accumulator were determined consideration of the convenient operation for operator on reloading. Boric acid concentrations of the RWST and the accumulators for Westinghouse 3-loop type plants were increased to meet the post-LOCA shutdown requirement for the long life cycles from 12 months to 18 months. To maintain LTC capability following a LOCA, the switchover time is examined using boron code of prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results showed that hot leg recirculation switchover times were shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3 and 4 and 8 hours from 18 hours for Ulchin 1 and 2, respectively. The flow path in the mode J for Kori 3 and 4 was recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1 and 2. (author). 2 tabs., 12 figs., 13 refs

  19. A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

    Directory of Open Access Journals (Sweden)

    Jae-Yong Kim

    2014-01-01

    Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

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

    International Nuclear Information System (INIS)

    Kim, Ho Kee

    1998-02-01

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

  1. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Perianez Alvarez, V.

    2013-07-01

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

  3. Analysis of cold leg LOCA with failed HPSI by means of integrated safety assessment methodology

    International Nuclear Information System (INIS)

    Gonzalez-Cadelo, J.; Queral, C.; Montero-Mayorga, J.

    2014-01-01

    Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  5. Radiative breaking of cosmologically acceptable grand unified theories

    International Nuclear Information System (INIS)

    Gato, B.; Leon, J.; Quiros, M.

    1984-01-01

    We present a cosmologically acceptable grand unified model where the breaking of SU(5) proceeds through radiative corrections induced by supergravity soft-breaking terms. The breaking scale is determined by dimensional transmutation. The model is compatible with the radiative breaking of SU(2)sub(L)xU(1)sub(Y) which provides an experimentally accessible low energy particle spectrum and small top quark mass. (orig.)

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

    International Nuclear Information System (INIS)

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

    1984-03-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

    2017-01-01

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

  8. Breaking the Paradox of Innovation in Small Businesses through Sustaining and Disruptive Reinvention

    Directory of Open Access Journals (Sweden)

    Vicki Baard

    2007-06-01

    Full Text Available In 2005 Deloitte Research released a paper examining the phenomenon they refer to asthe ‘innovation paradox’: the inability or reluctance of manufacturing firms to pursuestrategies that build the operational capabilities necessary for innovation that willprovide both profitability and growth. The report claims that this is due to the rapidlyincreasing complexity of global markets and the lack of synchronising innovation effortsacross their value chain, thus positioning the problem as an important contemporaryissue. While the research did not specifically target small and medium enterprises, theimplications for this business sector are considerable given their substantial contributionto global economies and their high failure rates in the first three to five years ofoperation. While not questioning the data in the Deloitte research, this paper doesquestion the assumption that the phenomenon is irreversible and the apparentunderlying self-fulfilling prophecy with respect to small to medium enterprises. Todemonstrate this the authors draw on a case study of a small manufacturing company inrural New South Wales, Australia, which operated between 1889 and 1983, to show thatthe breaking of the innovation paradox was successfully achieved by this firm in the latenineteenth and early twentieth century. Applying the case study to the Deloitte modelthe study demonstrates contemporary similarities by overlaying the Laycock history onthe successes / failures identified by Deloitte.

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  11. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

    1993-03-01

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

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

    International Nuclear Information System (INIS)

    Zeman, P.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  14. ROSA/LSTF experiment report for RUN SB-CL-24 repeated core heatup phenomena during 0.5% cold leg break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Mitsuhiro; Anoda, Yoshinari [Department of Reactor Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-03-01

    A small break loss-of-coolant accident (SBLOCA) in a Westinghouse-type four-loop PWR was simulated in an experiment (SB-CL-24) conducted at the Large-Scale Test Facility (LSTF) with an intention to study repeated core heatup during a long-term cooldown process. The experiment was conducted on February 28, 1990 with specified test conditions including failure assumptions both on the high pressure injection (HPI) and the auxiliary feedwater systems, and the intentional secondary system depressurization as an operator action. The secondary depressurization contributed to promote the primary depressurization and the actuation of accumulator injection system (AIS). A temporary core heatup was observed in each of three loopseal clearing (LSC) processes. A significant core heatup occurred in the following boil-off process after loss of the secondary coolant mass and the AIS termination due to increase of the primary pressure. By additional opening of the pressurizer relief valves and safety valves, the primary pressure rapidly decreased to result in the low pressure injection (LPI) which cooled the heated core. This report summarizes results of the experiment (SB-CL-24) in addition to typical responses of some accident indication systems including the core exit thermocouples (CETs) and the water level meters in the primary system. (author)

  15. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    International Nuclear Information System (INIS)

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

    1991-01-01

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B ampersand W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B ampersand W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions

  16. 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)

  17. Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA

    International Nuclear Information System (INIS)

    Cicero, G.M.; Briere, E.; Fornaciari, G.

    1994-06-01

    In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods

  18. A methodology for the estimation of release of fission products during LOCA with loss of ECCS

    International Nuclear Information System (INIS)

    Lele, H.G.; Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K.; Venkat Raj, V.

    2002-01-01

    A Loss of Coolant Accident (LOCA) in a nuclear reactor along with the failure of the Emergency Core Cooling System can cause sustained voiding of the core. In such a situation the core experiences very low flow which leads to poor heat removal from the reactor core. The heat to be removed from the core includes stored heat, heat generated due to metal water reaction at high temperatures, decay heat etc. The poor heat removal leads to heating of the fuel pins to high temperatures. The heating of fuel pins is further enhanced due to metal-water reaction at high temperatures. These high temperatures of the fuel pins can lead to fission product release, which is transported into the Primary Heat Transport (PHT) system and can enter the containment through the break. Analysis is involved due to the complexity of the system and the phenomena to be simulated. In this paper a multistage analysis methodology is presented that involves the development and application of a number of computer programs to model the various phenomena involved. The computer code PHTACT computes the activity release from the fuel as a function of fuel temperatures and cladding oxidation, its distribution into the PHT system and release into the containment. Computation of thermal hydraulic parameters during LOCA is done using the thermal hydraulic analysis code RELAP5. The detailed simulation of fuel pin temperatures is done using computer code HT/MOD4. The convective boundary conditions required for the code are obtained from RELAP5. Creep deformation is considered in the computation of dimensional changes of the coolant channel and estimation of flow blockage due to clad ballooning. The progression of various reaction layers due to high temperature reaction between fuel and clad and clad and steam is also computed, which affects the structural strength of the clad. Different approaches are possible and analysis can be carried out in different phases depending upon the complexities to be

  19. Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

    International Nuclear Information System (INIS)

    Bang, Young Seok

    2015-01-01

    The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage

  20. Analysis of Post-LOCA Core Inlet Blockage to Evaluate In-vessel Downstream Effect in APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    The method was developed to have a conservatism to cover the uncertainty of analysis and the acceptance is judged by the representative bounding estimation. However, the important safety parameters such as the available driving head need to be confirmed by the plant specific calculation. Also an interaction between the debris induced head loss and the core flow rate needs to be explained because the head loss induced by debris in actual condition may reduce the core inflow rate faster. To confirm the safety parameters, in this study, thermal-hydraulic response considering the core inlet blockage (CIB) by debris during LTCC process following a double-ended guillotine break of cold leg (CLB), one of hot leg (HLB) and one of intermediate leg (ILB) of the APR1400 were calculated, respectively. MARS-KS 1.3 code has been used. The CIB has been modeled by the closure of valves to the core in exponential manner with time to observe the behavior near the complete blockage. To understand the effect of core inlet blockage (CIB) during a long term core cooling (LTCC) phase following a loss-of-coolant accident (LOCA) in the light of in-vessel downstream effect (IDE) of Generic Safety Issue (GSI) 191, double-ended guillotine break of hot leg (HLB), one of cold leg (CLB) and one of intermediate leg (ILB) were calculated, respectively. And the important safety parameters such as the available driving head and the head loss due to debris were calculated using MARS-KS code and discussed in comparison with the WCAP method. As a result, a little delayed heatup behavior of the fuel cladding was found for all the cases, which due to the redistribution of flow within the core after blockage.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  2. Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

    In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

  3. Strongly coupled semidirect mediation of supersymmetry breaking

    International Nuclear Information System (INIS)

    Ibe, M.; Izawa, K.-I.; Nakai, Y.

    2009-01-01

    Strongly coupled semidirect gauge mediation models of supersymmetry breaking through massive mediators with standard-model charges are investigated by means of composite degrees of freedom. Sizable mediation is realized to generate the standard-model gaugino masses for a small mediator mass without breaking the standard-model symmetries.

  4. Development of 3-dimensional neutronics kinetics analysis code for CANDU-PHWR

    International Nuclear Information System (INIS)

    Kim, M. W.; Kim, C. H.; Hong, I. S.

    2005-02-01

    The followings are the major contents and scope of the research : development of kinetics power calculation module, formulation of space-dependent neutron transient analysis - implementation of 3-D and 2-G unified nodal method, verification of the kinetics module by benchmark problem - 3-D PHWR kinetics benchmark problem suggested by AECL, reactor trip simulation by shutdown system 1 in Wolsong unit 2. Development of a dynamic linked library code, SCAN D LL, for the coupled calculation with RELAP-CANDU : modeling of shutdown system 1, development of automatic shutdown module - automatic trip module based on rate log power control logic, automatic insertion of shutdown system 1. Development of a link code for coupled calculation - development of SCAN D LL(windows version), verification of coupled code by - 40% reactor inlet header break LOCA power pulse, 100% reactor outlet header break LOCA power pulse, 50% pump suction break LOCA power pulse

  5. Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1989-01-01

    This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

  6. Critical investigations and model development on countercurrent flow of gas and liquid in horizontal and vertical channels

    International Nuclear Information System (INIS)

    Mewes, D.; Beckmann, H.

    1989-01-01

    Countercurrent flow of steam and water occurs in the horizontal and vertical lines of a PWR in case of a LOCA. In order to predict the emergency core cooling behaviour in case of a large or small break LOCA it is important to calculate the volumetric flow rate of water which will get to the reactor core. Theoretical and experimental results of countercurrent flow in horizontal and vertical channels given by publication and reports are critically reviewed for the purpose of a more physical understanding of the flow phenomena. The influence of geometry, pressure and other boundary conditions are emphasized. The existing models which are developed to calculate the onset of flooding are based on experimental results of small test facilities. The applicability of these models to large geometries and high pressures as well as the consideration of condensation and entrainment are investigated. (orig./HP) [de

  7. Preparation and documentation of a CATHENA input file for Darlington NGS

    International Nuclear Information System (INIS)

    1989-03-01

    A CATHENA input model has been developed and documented for the heat transport system of the Darlington Nuclear Generating Station. CATHENA, an advanced two-fluid thermalhydraulic computer code, has been designed for analysis of postulated loss-of-coolant accidents (LOCA) and upset conditions in the CANDU system. This report describes the Darlington input model (or idealization), and gives representative results for a simulation of a small break at an inlet header

  8. Predictions of stratification in cold leg components using virtual noding schemes

    International Nuclear Information System (INIS)

    Piper, R.B.; Hassan, Y.A.; Banerjee, S.S.; Barsamian, H.R.; Cebull, P.P.

    1996-01-01

    In this investigation, a virtual noding scheme is used with RELAP5/MOD3.2 to capture thermal stratification effects in a small-break loss-of-coolant accident (LOCA) simulation. A three-dimensional code (CFD-ACE) has also been used to observe the stratification effects in a similar situation. Stratification temperature differences of the simulations compare well with that of the experiment. The Froude number was also evaluated

  9. Emergency response guide-B ECCS guideline evaluation analyses for N reactor

    International Nuclear Information System (INIS)

    Chapman, J.C.; Callow, R.A.

    1989-07-01

    INEL conducted two ECCS analyses for Westinghouse Hanford. Both analyses will assist in the evaluation of proposed changes to the N Reactor Emergency Response Guide-B (ERG-B) Emergency Core System (ECCS) guideline. The analyses were a sensitivity study for reduced-ECCS flow rates and a mechanistically determined confinement steam source for a delayed-ECCS LOCA sequence. The reduced-ECCS sensitivity study established the maximum allowable reduction in ECCS flow as a function of time after core refill for a large break loss-of-coolant accident (LOCA) sequence in the N Reactor. The maximum allowable ECCS flow reduction is defined as the maximum flow reduction for which ECCS continues to provide adequate core cooling. The delayed-ECCS analysis established the liquid and steam break flows and enthalpies during the reflood of a hot core following a delayed ECCS injection LOCA sequence. A simulation of a large, hot leg manifold break with a seven-minute ECCS injection delay was used as a representative LOCA sequence. Both analyses were perform using the RELAP5/MOD2.5 transient computer code. 13 refs., 17 figs., 3 tabs

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    CSNI therefore posed the question to the Working Group on Fuel Safety (WGFS): How could the Halden LOCA tests affect regulation? The WGFS agreed that the main safety concern would be fuel dispersal (and hence the potential for loss of coolable geometry) occurring at relatively low temperature, i.e. 800 deg. C. In order to assess the applicability of the IFA-650.4 results to actual power plant situations and the possible impact on safety criteria, a number of aspects should be clarified before considering a safety significance of the Halden IFA-650 series results: - Representativeness for NPP cases - Gas flow - Relocation - Burnup effect - Repeatability - Power history These items will be discussed one by one in this CSNI report. On April 17, 2009, test 650.9 was carried out with 650.4 sibling fuel. The target cladding peak temperature was 1100 deg. C in this case, but otherwise the experimental conditions were very similar. In many respects, 650.9 repeated the 650.4 experiment, e.g. by showing clear signs of fuel relocation which was confirmed by gamma scanning later on. The WGFS therefore decided that 650.9 should be considered as well for this CSNI report. Mention is also made of IFA-650.3, which failed with a small crack in a weak spot induced by thermocouple welding, and IFA-650.5 which involved ballooning and fuel ejection under conditions of restricted gas flow

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

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1986-01-01

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

  13. Analysis of two small break loss-of-coolant experiments in the BETHSY facility using RELAP5/MOD3

    International Nuclear Information System (INIS)

    Roth, P.A.; Schultz, R.R.; Choi, C.J.

    1992-07-01

    Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system depressurization, break flow rate, core heat-up, and effect of operator actions were analyzed. Based on the test 9.lb/ISP-27 and 6.2 TC data, an assessment study of the RELAP5/MOD3 version 7 code was performed which included a study of the above phenomena along with countercurrent flow limitation and vapor pull-through. The code provided a reasonable simulation of the various phenomena which occurred during the tests

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    Science.gov (United States)

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

    2018-02-01

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

  16. 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.

  17. 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

  18. 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.

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

    International Nuclear Information System (INIS)

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

    1982-09-01

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

  20. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  1. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    International Nuclear Information System (INIS)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6'' cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author)

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

    International Nuclear Information System (INIS)

    Perianez Alvarez, V.

    2013-01-01

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

  3. Comparative analysis of SLB for OPR1000 by using MEDUSA and CESEC-III codes

    International Nuclear Information System (INIS)

    Park, Jong Cheol; Park, Chan Eok; Kim, Shin Whan

    2005-01-01

    The MEDUSA is a system thermal hydraulics code developed by Korea Power Engineering Company (KOPEC) for Non-LOCA and LOCA analysis, using two fluid, three-field governing equations for two phase flow. The detailed descriptions for the MEDUSA code are given in Reference. A lot of effort is now being made to investigate the applicability of the MEDUSA code especially to Non-LOCA analysis, by comparing the analysis results with those from the current licensing code, CESEC-III: The comparative simulations of Pressurizer Level Control System(PLCS) Malfunction and Feedwater Line Break(FLB), which have been accomplished by C.E.Park and M.T.Oh, respectively, already showed that the MEDUSA code is applicable to the analysis of Non-LOCA events. In this paper, detailed thermal hydraulic analyses for Steam Line Break(SLB) without loss of off-site power were performed using the MEDUSA code. The calculation results were also compared with the CESEC-III, 1000(OPR1000), for the purpose of the code verification

  4. Yukawa unification in moduli-dominant SUSY breaking

    International Nuclear Information System (INIS)

    Khalil, S.; Tatsuo Kobayashi

    1997-07-01

    We study Yukawa in string models with moduli-dominant SUSY breaking. This type of SUSY breaking in general leads to non-universal soft masses, i.e. soft scalar masses and gaugino masses. Such non-universality is important for phenomenological aspects of Yukawa unification, i.e., successful electroweak breaking, SUSY corrections to the bottom mass and the branching ratio of b → sγ. We show three regions in the whole parameter space which lead to successful electroweak breaking and allow small SUSY corrections to the bottom mass. For these three regions we investigated the b → sγ decay and mass spectra. (author). 26 refs, 6 figs

  5. Comparison of models discribing cladding deformations during LOCA

    International Nuclear Information System (INIS)

    Chakraborty, A.K.; Zipper, R.

    1981-05-01

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

  6. 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)

  7. 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)

  8. 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.

  9. Development of the MARS input model for Kori nuclear units 1 transient analyzer

    International Nuclear Information System (INIS)

    Hwang, M.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Jeong, J. J.

    2004-11-01

    KAERI has been developing the 'NSSS transient analyzer' based on best-estimate codes for Kori Nuclear Units 1 plants. The MARS and RETRAN codes have been used as the best-estimate codes for the NSSS transient analyzer. Among these codes, the MARS code is adopted for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. So it is necessary to develop the MARS input model for Kori Nuclear Units 1 plants. This report includes the input model (hydrodynamic component and heat structure models) requirements and the calculation note for the MARS input data generation for Kori Nuclear Units 1 plant analyzer (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Kori Nuclear Units 1

  10. RELAP5 - a new tool for pressurized water reactor safety analysis

    International Nuclear Information System (INIS)

    Perneczky, L.

    1988-11-01

    The RELAP type pressurized water reactor safety system codes are used world wide for the loss of coolant accident analyses. In this paper the RELAP5, the advanced generation of the code family is presented. The relationship to RELAP4/mod6 version is discussed. The capability of the RELAP5/mod1-EUR version for small, medium and large break LOCA is investigated based on international user experience. (author) 30 refs

  11. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  12. 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

  13. LOCA testing of high burnup PWR fuel in the HBWR. Additional PIE on the cladding of the segment 650-5

    Energy Technology Data Exchange (ETDEWEB)

    Oberlaender, B.C.; Espeland, M.; Jenssen, H.K.

    2008-07-01

    IFA-650.5, a test with pre-irradiated fuel in the Halden Project LOCA test series, was conducted on October 23rd, 2006. The fuel rod had been used in a commercial PWR and had a high burnup, 83 MWd/kgU. Experimental arrangements of the fifth test were similar to the preceding LOCA tests. The peak cladding temperature (PCT) level was higher than in the third and fourth tests, 1050 C. A peak temperature close to the target was achieved and cladding burst occurred at approx. 750 C. Within the joint programme framework of the Halden Project PIE was done, consisting of gamma scanning, visual inspection, neutron-radiography, hydrogen analysis and metallography / ceramography. An additional extensive PIE including metallography, hydrogen analysis, and hardness measurements of cross-sections at seven axial elevations was done. It was completed to study the high burnup and LOCA induced effects on the Zr-4 cladding, namely the migration of oxygen into the cladding from the inside surface, the cladding distension, and the burst (author)(tk)

  14. Staff discussion of fifteen technical issues listed in attachment to November 3, 1976 memorandum from Director, NRR to NRR staff

    International Nuclear Information System (INIS)

    1976-11-01

    Issues discussed include: treatment of non-safety grade equipment in evaluation of a postulated steamline break accidents, lack of independence of interlocks in ECCS valves, acceptability of swing bus design of BWR-4 plants, loss of offsite power subsequent to manual safety injection reset following a LOCA, postulated reactor coolant pump rotor seizures, single failures in reactivity control, passive failures following a LOCA, probabilistic assessment of reliability, frequency decay, grid stability, interpretation of GDC 19, load break switch, instrument trip setpoints, computer protection system, and overpressurization

  15. ROSA-II test data report, 10

    International Nuclear Information System (INIS)

    1977-12-01

    Results of the ROSA-II test simulating a loss-of coolant accident (LOCA) in a light water reactor (LWR) are presented, including the test conditions and interpretation of the phenomena for test runs 415, 417, 421 and 422. Even in small break at the cold leg, the core is exposed to void and the temperature rises. In small break of the hot leg, however, core cooling keeps without temperature rise, because there still remains much residual water and upward core flow exists. Direct effect of the HPCI on the depressurization rate is small, but it increases the accumulator injection rate, leading to early core reflooding and early core cooling from upward. Effects of the secondary system depressurization are increase of depressurization and discharge rates of the primary loop, which results in early initiation of the accumulator injection and core reflooding. (auth.)

  16. 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)

  17. A study on effective system depressurization during a PWR vessel bottom break LOCA with HPI failure and gas inflow prevention. ROSA-V/LSTF test SB-PV-05

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2006-11-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-05, simulated a PWR vessel bottom SBLOCA with a rupture of nine instrument tubes, which is equivalent to 0.18% cold leg break. It is clarified that AM actions with steam generator (SG) depressurization to achieve a primary loop cooling rate at -55 K/h and auxiliary feedwater supply for 30 minutes are effective to avoid core uncovery by actuating the low pressure injection (LPI) system. It is also shown through the comparison with the previous experiment of SB-PV-03 that prevention of non-condensable gas inflow from the accumulator injection system (AIS) is very important to actuate the LPI to achieve adequate core cooling. This report presents experiment results of SB-PV-05 in detail and shows the effects of gas inflow prevention on core cooling through the estimation of primary coolant mass and energy balance in the primary system. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  1. Lifshitz-sector mediated SUSY breaking

    International Nuclear Information System (INIS)

    Pospelov, Maxim; Tamarit, Carlos

    2014-01-01

    We propose a novel mechanism of SUSY breaking by coupling a Lorentz-invariant supersymmetric matter sector to non-supersymmetric gravitational interactions with Lifshitz scaling. The improved UV properties of Lifshitz propagators moderate the otherwise uncontrollable ultraviolet divergences induced by gravitational loops. This ensures that both the amount of induced Lorentz violation and SUSY breaking in the matter sector are controlled by Λ HL 2 /M P 2 , the ratio of the Hořava-Lifshitz cross-over scale Λ HL to the Planck scale M P . This ratio can be kept very small, providing a novel way of explicitly breaking supersymmetry without reintroducing fine-tuning. We illustrate our idea by considering a model of scalar gravity with Hořava-Lifshitz scaling coupled to a supersymmetric Wess-Zumino matter sector, in which we compute the two-loop SUSY breaking corrections to the masses of the light scalars due to the gravitational interactions and the heavy fields

  2. Development of gamma-ray densitometer and measurement of void fraction in instantaneous pipe rupture under BWR LOCA condition

    International Nuclear Information System (INIS)

    Yano, Toshikazu

    1983-11-01

    In order to clarify the transient mass flow rate under the instantaneous pipe rupture condition, it is necessary to use a highly sensitive void meter. Therefore, a high-response gamma-ray densitometer was developed for the measurement of void fraction variation caused by flashing vaporization of the high-pressure and -temperature water under the instantaneous pipe rupture accident. The measurement of void fraction was performed in the pipe rupture test under the BWR LOCA condition with a 6-inch diameter pipe. Initial conditions of the water were 6.86 MPa in pressure and the saturation temperature. To prove the reliability and accuracy, a calibration test by falling acrylic void simulators and an air injection test into cold water filled in the pipe were also conducted. The following results are obtained in the pipe rupture test. (1) The cone slit method is very useful to increase the measuring accuracy. (2) It is clearly observed that the apparent increase of void fraction occurs after the rarefaction wave passes. (3) The first maximum of void fraction occurs with some delay time after break. The following minimum void fraction concurs with the maximum pressure in the pressure recovering phenomena and with the maximum blowdown thrust force. (author)

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  4. Simulation of long-term cooling in the VVER-640 power plant after a large break LOKA on the PACTEL facility

    International Nuclear Information System (INIS)

    Banati, J.

    2000-01-01

    The present report gives a short introduction to the safety features of the new Russian VVER-640 reactor design. In order to analyze the complex thermal hydraulic phenomena during long-term cooling after a large-break LOCA, experiments will be carried out in the PACTEL facility. For preparation, pre-test calculations were performed using the RELAPS/MOD3.2 computer code. The main part of the report discusses the results obtained by the program. The structure and options used in the input deck, as well as the efforts of code application to the simulation of proposed experiments are reviewed. A short sensitivity study is provided on the calculated results. Finally, conclusions are drawn for the code capabilities to represent the expectable trends in the upcoming tests. (orig.)

  5. Gravitino dark matter in R-parity breaking vacua

    International Nuclear Information System (INIS)

    Buchmueller, W.; Covi, L.; Ibarra, A.; Hamaguchi, K.; Yanagida, T.T.

    2007-02-01

    We show that in the case of small R-parity and lepton number breaking couplings, primordial nucleosynthesis, thermal leptogenesis and gravitino dark matter are naturally consistent for gravitino masses m 3/2 >or similar 5 GeV. We present a model where R-parity breaking is tied to B-L breaking, which predicts the needed small couplings. The metastable next-to-lightest superparticle has a decay length that is typically larger than a few centimeters, with characteristic signatures at the LHC. The photon flux produced by relic gravitino decays may be part of the apparent excess in the extragalactic diffuse gamma-ray flux obtained from the EGRET data for a gravitino mass m 3/2 ∝10 GeV. In this case, a clear signal can be expected from GLAST in the near future. (orig.)

  6. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  7. Development, assessment, licensing, and application of RELAP5YA at Yankee Atomic Electric Company

    International Nuclear Information System (INIS)

    Husain, A.

    1984-01-01

    Since 1975, Yankee Atomic Electric Company (YAEC) has been licensed to perform LOCA analyses (large and small breaks) for PWRs. The methods are based upon the WREM package, comprised of RELAP4-MOD 3, RELAP4-FLOOD, and TOODEE-2EM. Significant efforts were expended to compare WREM models against Appendix K criteria, and to incorporate changes for compliance with the criteria and for improved representation of LOCA phenomena. /SUP 2,3,4/ This internalized capability has been used extensively for licensing the Yankee and Maine Yankee plants, and for timely resolution of LOCA issues which frequently arose. The value of this internalized analysis capability was well recognized by the Yankee management. In 1979, Yankee embarked upon another LOCA Methods Development Program. One objective was to remove extra conservatisms inherent in older codes to provide additional operational flexibility for our plants. Also, we needed a code which could be used in a production mode to provide realistic response for off-normal conditions to be used in operator training and emergency guideline assessment. This program resulted in the development of RELAP5YA. The development, assessment, licensing, and application of RELAP5YA at YAEC is briefly described in this paper

  8. 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

  9. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  10. LOFA [loss of flow accident] and LOCA [loss of coolant accident] in the TIBER-II engineering test reactor: Appendix A-4

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510 0 C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs

  11. ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R; Shen, K; Sjoeberg, A

    1995-03-01

    This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs.

  12. ITER pressure and thermal loads to containment HTS vault LOCA analysis. Draft final report EC Task SEA 3, Subtask 3-4

    International Nuclear Information System (INIS)

    Blomquist, R.; Shen, K.; Sjoeberg, A.

    1995-03-01

    This study has been performed within the framework of the EC Task SEA 3 and its objective is to provide necessary data in supporting the design solution of the ITER secondary confinement around the primary heat transfer system equipment. These data relate to the required dimensions for the blow-out panels, the vent lines and the suppression tank following a LOCA in one of the HTS vaults, namely the first wall/shielding blanket(FW/SB) vault, divertor vault and vacuum vessel (VV) vault. In this report, we present the design and operational input and describe the identified accident sequences. The input data are in correspondence with ITER design data of November 1994. The computer codes used are RELAP5 (LOCA flows) and CONTAIN (secondary confinement thermal-hydraulics) and models of calculations are given. The results in the form of diagrams demonstrating transients of various variables after a LOCA, are presented. After some discussions of the results, we indicate some topics for the continuing study with the emphasis on optimization of the containment system. 10 refs, 29 figs

  13. Climate and floods still govern California levee breaks

    Science.gov (United States)

    Florsheim, J.L.; Dettinger, M.D.

    2007-01-01

    Even in heavily engineered river systems, climate still governs flood variability and thus still drives many levee breaks and geomorphic changes. We assemble a 155-year record of levee breaks for a major California river system to find that breaks occurred in 25% of years during the 20th Century. A relation between levee breaks and river discharge is present that sets a discharge threshold above which most levee breaks occurred. That threshold corresponds to small floods with recurrence intervals of ???2-3 years. Statistical analysis illustrates that levee breaks and peak discharges cycle (broadly) on a 12-15 year time scale, in time with warm-wet storm patterns in California, but more slowly or more quickly than ENSO and PDO climate phenomena, respectively. Notably, these variations and thresholds persist through the 20th Century, suggesting that historical flood-control effects have not reduced the occurrence or frequency of levee breaks. Copyright 2007 by the American Geophysical Union.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

    Energy Technology Data Exchange (ETDEWEB)

    Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

    1995-09-01

    The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

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

    International Nuclear Information System (INIS)

    Mattila, L.J.

    1977-01-01

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

  17. Probabilistic safety assessment of the nuclear facilities in Cuba

    International Nuclear Information System (INIS)

    Rivero O, J.J.; Salomon L, J.

    1991-01-01

    During 1986-1990 basis were established for further developing probabilistic safety assessment (PSA) of Juragua NPP. A team work was consolidated and carried out the preliminary studies of the small break LOCA initiating event. A significant achievement was the creation of the ANCON code, which allows the evaluation of complex fault trees in personal computers, and has been applied in PSA modelling, and specialist qualification. The paper describes the main results and future activities in this field. (author)

  18. Gravitino dark matter in R-parity breaking vacua

    Energy Technology Data Exchange (ETDEWEB)

    Buchmueller, W.; Covi, L.; Ibarra, A. [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Hamaguchi, K.; Yanagida, T.T. [Tokyo Univ. (Japan). Dept. of Physics

    2007-02-15

    We show that in the case of small R-parity and lepton number breaking couplings, primordial nucleosynthesis, thermal leptogenesis and gravitino dark matter are naturally consistent for gravitino masses m{sub 3/2} >or similar 5 GeV. We present a model where R-parity breaking is tied to B-L breaking, which predicts the needed small couplings. The metastable next-to-lightest superparticle has a decay length that is typically larger than a few centimeters, with characteristic signatures at the LHC. The photon flux produced by relic gravitino decays may be part of the apparent excess in the extragalactic diffuse gamma-ray flux obtained from the EGRET data for a gravitino mass m{sub 3/2}{proportional_to}10 GeV. In this case, a clear signal can be expected from GLAST in the near future. (orig.)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

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