scholarly journals ROSA/LSTF Tests and Posttest Analyses by RELAP5 Code for Accident Management Measures during PWR Station Blackout Transient with Loss of Primary Coolant and Gas Inflow

2018 ◽  
Vol 2018 ◽  
pp. 1-19 ◽  
Author(s):  
Takeshi Takeda ◽  
Iwao Ohtsu
Keyword(s):  
Flow Rate ◽  
Injection System ◽  
Test Facility ◽  
Relief Valve ◽  
Nitrogen Gas ◽  
Primary Coolant ◽  
Gas Accumulation ◽  
Accident Management ◽  
Station Blackout ◽  
Secondary Side

Three tests were carried out with the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated accident management (AM) measures during station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total failure of high-pressure injection system in a pressurized water reactor. As the AM measures, steam generator (SG) secondary-side depressurization was done by fully opening the relief valves in both SGs, and auxiliary feedwater was injected into the secondary-side of both SGs simultaneously. Conditions for the break size and the onset timing of the AM measures were different among the three LSTF tests. In the three LSTF tests, the primary pressure decreased to a certain low pressure of below 1 MPa with or without the primary depressurization by fully opening the relief valve in a pressurizer as an optional AM measure, while no core uncovery took place through the whole transient. Nonuniform flow behaviors were observed in the SG U-tubes under natural circulation (NC) with nitrogen gas depending probably on the gas accumulation rate in the two LSTF tests that the gas accumulated remarkably. The RELAP5/MOD3.3 code predicted most of the overall trends of the major thermal hydraulic responses observed in the three LSTF tests. The code, however, indicated remaining problems in the predictions of the primary pressure, the SG U-tube collapsed liquid levels, and the NC mass flow rate after the nitrogen gas ingress as well as the accumulator flow rate through the analyses for the two LSTF tests, where the remarkable gas accumulation occurred.

scholarly journals ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

2016 ◽  
Vol 2016 ◽  
pp. 1-15
Author(s):  
Takeshi Takeda ◽  
Akira Ohnuki ◽  
Daisuke Kanamori ◽  
Iwao Ohtsu
Keyword(s):  
Steam Generator ◽  
Safety System ◽  
Injection System ◽  
Test Facility ◽  
Gas Release ◽  
Dissolved Gas ◽  
Nitrogen Gas ◽  
Two Phase ◽  
Pressurized Water ◽  
Secondary Side

Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

ROSA/LSTF Experiment on a PWR Station Blackout Transient With AM Measures and RELAP5 Post-Test Analysis

2014 ◽  
Author(s):  
Takeshi Takeda ◽  
Iwao Ohtsu ◽  
Taisuke Yonomoto
Keyword(s):  
Flow Rate ◽  
Large Scale ◽  
Heat Removal ◽  
Sensitivity Analyses ◽  
Test Facility ◽  
Primary System ◽  
Energy Agency ◽  
Coolant Injection ◽  
Station Blackout ◽  
Secondary Side

An experiment on a PWR station blackout transient with the TMLB’ scenario and accident management (AM) measures was conducted using the ROSA/large scale test facility (LSTF) at Japan Atomic Energy Agency under an assumption of non-condensable gas inflow to the primary system from accumulator (ACC) tanks. The AM measures proposed in this study are steam generator (SG) secondary-side depressurization by fully opening the safety valves in both SGs with the start of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The LSTF test revealed the primary pressure started to decrease when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after the completion of ACC coolant injection. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of SG U-tube collapsed liquid level and primary mass flow rate after the gas ingress. The SG coolant injection flow rate was found to affect significantly the peak cladding temperature and the ACC actuation time through the RELAP5 sensitivity analyses.

RELAP5 Code Uncertainty Analysis of LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization as an Accident Management Procedure

2016 ◽  
Author(s):  
Ikuo Kinoshita ◽  
Toshihide Torige
Keyword(s):  
Steam Generator ◽  
Injection System ◽  
Test Facility ◽  
Test Cases ◽  
Input Uncertainty ◽  
Accident Scenario ◽  
Management Procedure ◽  
Accident Management ◽  
Uncertainty Parameters ◽  
Secondary Side

The Best Estimate Plus Uncertainty (BEPU) method is applied to analysis of the “intentional depressurization of steam generator secondary side” which is an accident management procedure in a small break loss-of-coolant accident with high pressure injection system failure. In the present study, experimental analyses using the RELAP5/MOD3.2 code were carried out for the ROSA/Large Scale Test Facility (LSTF) secondary-side depressurization tests. The two test cases were selected with different break sizes and different depressurization conditions to ensure the reliability for the accident scenario analyses. The input parameter uncertainty propagation analyses were performed to get 95%/95% tolerance limit values of the output parameters. It was confirmed that the code predicted well the major event progressions of the accident for both test cases and the 95%/95% uncertainty bounds of the peak cladding temperatures included the measured values. On the other hand, the same ranges of some input uncertainty parameters could lead to different influences on the output uncertainties between the test cases. The dominating input uncertainty parameters could be different depending on the break sizes and depressurization conditions of the accident scenario.

RELAP5 Code Analysis of LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization as an Accident Management Procedure: Investigation on Base Case Result Appropriate for the Best Estimate Plus Uncertainty (BEPU) Application

2014 ◽  
Author(s):  
Ikuo Kinoshita ◽  
Toshihide Torige ◽  
Michio Murase ◽  
Yoshitaka Yoshida ◽  
Takeshi Takeda ◽  
...  
Keyword(s):  
Steam Generator ◽  
Large Scale ◽  
Model Simulation ◽  
Test Facility ◽  
Base Case ◽  
Best Estimate ◽  
Management Procedure ◽  
Accident Management ◽  
Characteristic Features ◽  
Secondary Side

The application of the Best Estimate Plus Uncertainty (BEPU) method is made to analysis of the “Intentional depressurization of steam generator secondary side” which is an accident management procedure in a small-break loss-of-coolant accident (SBLOCA) with high pressure injection (HPI) system failure. RELAP5/MOD3.2 is used as the analysis code. By applying the BEPU method, the uncertainties of the analysis results can be estimated quantitatively. However, the accuracy of the analysis results depends primarily on the base case result predicted by the best estimate code. In this study, in order to investigate the appropriate base case model, simulation analyses using the RELAP5/MOD3.2 were carried out for the ROSA Large Scale Test Facility (ROSA/LSTF) secondary-side depressurization tests. It was found that the code predicted well the major event progressions such as pressure responses, core liquid level responses, and rod surface temperatures, as well as important phenomena such as formation and clearing of loop seals, accumulation of water from condensation, and countercurrent flow limitation (CCFL) at the inlet of the U-tubes, which are characteristic features of this accident scenario.

Integral Effect Test for Performance Evaluation of the PAFS (Passive Auxiliary Feedwater System) During a SBO (Station Blackout) Transient

2014 ◽  
Author(s):  
Byoung-Uhn Bae ◽  
Seok Kim ◽  
Yu-Sun Park ◽  
Yun-Je Cho ◽  
Kyoung-Ho Kang
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Water Levels ◽  
Experimental Result ◽  
Test Facility ◽  
Decay Heat ◽  
The Core ◽  
Integral Effect ◽  
Station Blackout ◽  
Effect Test

Station blackout (SBO) accident is considered as one of the most significant design extension conditions (DECs), which has been extensively focused after the Fukushima Dai-chi accident. When the SBO accident occurs in the APR+ (Advance Power Reactor Plus), the PAFS (Passive Auxiliary Feedwater System), which is an advanced safety feature adopted in the APR+, should play a significant role to cool down the core decay heat without any operation of active safety systems. This study focuses on validation of the cooling and operational performance for the PAFS during the SBO transient with utilizing an integral effect test facility, ATLAS-PAFS. In order to simulate the SBO transient of the APR+ as realistically as possible, a pertinent scaling approach was taken into account. The initial steady-state conditions and the sequence of event in the SBO scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. In the transient simulation, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate at the return-water line were measured and investigated. Following the reactor trip at the initiation of the transient, the coolant inventory of the secondary system of the steam generator was reduced by the repeated opening and closing of the MSSV. When the collapsed water level reached 25% of wide range, the PAFS was actuated to cool down the primary system by the condensation heat transfer at the PCHX (Passive Condensation Heat Exchanger). The pressure and the temperature of the reactor coolant system continuously decreased during the heat removal by the PAFS operation. It points out that the PAFS can supply auxiliary feedwater to the steam generator and remove the core decay heat without any active system. From the present experimental result, it could be concluded that the APR+ has the capability of coping with the hypothetical SBO scenario with adopting the PAFS and proper set-points of its operation. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS, RELAP5 as well as the SPACE code and to identify any code deficiency for a SBO simulation with an operation of the PAFS.

RELAP5 Code Analysis of LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization and its Uncertainty Quantification by Monte Carlo Method and Wilks Formula Approach

2016 ◽  
Author(s):  
Ikuo Kinoshita ◽  
Michio Murase
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Steam Generator ◽  
Tolerance Limit ◽  
Injection System ◽  
Test Facility ◽  
Statistical Fluctuation ◽  
Limit Values ◽  
Upper Limit ◽  
Secondary Side

The Best Estimate Plus Uncertainty (BEPU) method has been applied by the authors to analysis of the “intentional depressurization of steam generator secondary side” which is an accident management procedure in a small break loss-of-coolant accident with high pressure injection system failure. In the present study, experimental analyses using the RELAP5/MOD3.2 code were carried out for the ROSA/Large Scale Test Facility (LSTF) secondary-side depressurization tests. The two test cases were selected with different break sizes and different depressurization conditions to ensure the reliability for the accident scenario analyses. The uncertainty propagation analyses were performed through the random variations of input parameters whose uncertainty ranges and distributions were determined previously by the PIRT and the separate effects tests. One thousand random calculations were conducted to get the 95% upper limit values of the peak cladding temperature (PCT) by the Monte Carlo method. Furthermore, the 95%/95% tolerance limits for the PCT were obtained according to Wilks formula. It was confirmed that the code predicted well the major event progressions such as rod surface temperature and the 95% uncertainty bands included the measured values. Furthermore, the 95% upper limit values of the PCT are below the 95%/95% tolerance limit values. However, the statistical fluctuation of the tolerance limit values by Wilks first order formula is as large as the uncertainty value itself. The statistical fluctuation decreases with increasing order of Wilk formula. It is desirable to increase the order of Wilks formula to more than the second order to get the reliable safety margin.

Assessment of Early-Phase Accident Management Strategies in a Station Blackout Scenario for VVER-1000 Reactors

2010 ◽  
Author(s):  
Polina Tusheva ◽  
Frank Schaefer ◽  
Nils Reinke ◽  
Frank-Peter Weiss
Keyword(s):  
High Pressure ◽  
Early Phase ◽  
Management Strategies ◽  
Feed Water ◽  
Water Tank ◽  
Severe Accidents ◽  
Accident Management ◽  
Station Blackout ◽  
Safety Systems ◽  
Secondary Side

In recent years, many NPPs have developed and implemented severe accident management guidelines (SAMG). It is the primary objective of developing SAMG to prevent or mitigate the consequences of severe accidents by keeping the reactor pressure vessel (RPV) integrity and reducing the load to the containment. In a hypothetical Station Blackout accident all active safety systems are unavailable. Without additional measures this would lead to heating-up of the reactor core with severe core degradation. To avoid or to limit the consequences of a possible core heat up, different accident management strategies can be applied. This paper presents an assessment of early-phase accident management actions for VVER-1000 reactors. In particular Primary Side Depressurization (PSD) is investigated as a basic strategy for managing severe accidents under high pressure conditions. In addition, Secondary Side Depressurization (SSD) is also being investigated. It aims at fast reduction of the secondary pressure and feeding the steam generators’ secondary side with water from the feed water tank or from a different source. In that way, the heat removal from the primary to the secondary side can be significantly enhanced and the core heat-up at high pressure can be delayed. A number of simulations with different criteria for actuation of the PSD procedure and additional SSD were performed using the thermal-hydraulic system code ATHLET. This paper provides a detailed modelling of the reactor coolant system and the required safety systems, analysis of the thermal-hydraulic and safety parameters and description of the physical phenomena. Special attention is given to the possibilities of preventing or at least delaying an extended core heat-up depending on the availability of the operational and safety systems. The effectiveness of the applied accident management measures and the effect on the accident progression were studied in order to assess the maximum response time for operators’ intervention.

scholarly journals RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation

2012 ◽  
Vol 2012 ◽  
pp. 1-15 ◽  
Author(s):  
Takeshi Takeda ◽  
Hideaki Asaka ◽  
Hideo Nakamura
Keyword(s):  
Flow Rate ◽  
High Power ◽  
Natural Circulation ◽  
Liquid Level ◽  
Injection System ◽  
Relief Valve ◽  
Two Phase ◽  
Fine Mesh ◽  
Post Test ◽  
Long Time

A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater (LOFW) transient with specific assumptions of failure of scram that may cause natural circulation with high core power and total failure of high pressure injection system. Auxiliary feedwater (AFW) was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and steam generator (SG) secondary-side pressures were maintained, respectively, at around 16 and 8 MPa by cycle opening of pressurizer (PZR) power-operated relief valve and SG relief valves for a long time. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed to well understand the observed phenomena by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis counter-current flow limiting correlation at the inlet of U-tubes. The code, however, has remaining problems in proper predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level.

Some Lessons Learned From the SIPACT Simulations on the Design of PWR and Improvement of AM Measures

2002 ◽  
Author(s):  
R. Pochard ◽  
F. Jedrzejewski ◽  
S. Nilsuwankosit
Keyword(s):  
Pressure Vessel ◽  
Bleeding Time ◽  
Total Loss ◽  
Injection System ◽  
Feed Water ◽  
Relief Valve ◽  
Pressurized Water ◽  
Primary Loop ◽  
The Core ◽  
Secondary Side

In the general context of the nuclear activities, life extension of the existing plants is the interesting option for countries that are already well equipped with NPPs. As the working life of 60 years is now expected possible for some well maintained plants, their safety measures needs to be improved such that they should be comparable to the new or future designs, taken into account the results from the probabilistic and the deterministic accident analysis. To accomplish this aim, the Accident Management (AM) is the important part of the process that must be utilized including possible automation of some processes. At INSTN, the extensive sensitivity studies related to the feed and bleed process on the primary and the secondary side had been carried out with the SIPACT simulator, based on the Cathare code, for a 900 MWe pressurized water reactor. The simulations had been mainly conducted for the Beyond Design Basis Accident (BDBA) condition. This condition included the total loss of feed-water and a small break with the loss of the high pressure injection system (HPIS). From these studies, several interesting findings had been obtained. For AM purpose and with the bleeding process, the criterion called “the safety time margin” for core uncovery was introduced. By plotting the safety time margin against the bleeding time, the relation between them was established and used to optimize, when possible, the AM measures. For the scenario that involved the total loss of feed water, in case of full bleeding, a window was found for the bleeding time around the degradation of the heat exchange in SGs would be resulted. In this scenario, one of the solutions was to open only one relief valve at first in order to let through only the minimal mass. At the time of the injection by the accumulator, the other two relief valves were then opened. As a result, the flow through the relief valves could be effectively compensated by the flow from the accumulator, the mass balance in the vessel was maintained and the safety margin time was increased. For the scenario that was related to a small break without HPIS, the concept of the safety time margin was still applicable. The time window was observed to be narrower for the bleeding on the secondary side if the core uncovery was to be avoided, however. By observing the distribution of the mass in the primary loop, its behavior, which was directly related to the design, was fully demonstrated. One important finding showed that the current PWR design presented some disadvantage under the BDBA condition. Due to the way the water was accumulated in various components, sometime as much as that that was still remained in the pressure vessel, not all the water already presented or injected into the primary loop could reach the pressure vessel to be effectively utilized for core cooling. In order to characterize the availability of the water to cool the core, which related to the NPP BDBA robustness, a simple mass distribution criterion was proposed. Some improvements for the future design were also suggested.

Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

2012 ◽  
Vol 250 ◽  
pp. 633-645 ◽  
Author(s):  
A. Bucalossi ◽  
A. Del Nevo ◽  
F. Moretti ◽  
F. D’Auria ◽  
I.V. Elkin ◽  
...  
Keyword(s):  
Test Facility ◽  
Integral Test ◽  
Management Procedures ◽  
Accident Management ◽  
Station Blackout
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