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.

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.

Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

2008 ◽  
Author(s):  
Mitsuhiro Suzuki ◽  
Takeshi Takeda ◽  
Hideo Nakamura
Keyword(s):  
Time Delay ◽  
Large Scale ◽  
Simulation Experiment ◽  
Three Dimensional ◽  
Test Facility ◽  
Large Temperature ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Accident Management

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 sides in case of core temperature excursion. Test 6-1 is the 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 reasons 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.

Analysis of the Core Exit Temperature and the Peak Cladding Temperature During a SBLOCA: Application to a Scaled-Up Model

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Andrea Querol ◽  
Sergio Gallardo ◽  
Gumersindo Verdú
Keyword(s):  
Time Delay ◽  
Large Scale ◽  
Test Facility ◽  
The Core ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Exit Temperature ◽  
Scale Test ◽  
Versus Peak ◽  
Physical Phenomena

During loss-of-coolant accidents (LOCAs), operators may start accident management (AM) actions when the core exit temperature (CET) measured by thermocouples exceeds a certain value. However, a significant time delay and temperature discrepancy in the superheat detection were observed in several facilities. This work is focused on clarifying CET thermocouple responses versus peak cladding temperature (PCT) and studying if the same physical phenomena are reproduced in two TRACE5 models with different geometry (a large-scale test facility (LSTF) and a scaled-up LSTF) during a pressure vessel (PV) upper head small break LOCA (SBLOCA). Results obtained show that the delay between the core uncover and the CET excursion is reproduced in both cases.

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.

Sampling Based Uncertainty Analysis of 10% Hot Leg Break LOCA in LSTF

2010 ◽  
Author(s):  
Samiran Sengupta ◽  
S. K. Dubey ◽  
R. S. Rao ◽  
S. K. Gupta ◽  
V. K. Raina
Keyword(s):  
Sensitivity Analysis ◽  
Uncertainty Analysis ◽  
Large Scale ◽  
Latin Hypercube Sampling ◽  
Test Facility ◽  
Accident Analysis ◽  
Design Matrix ◽  
Best Estimate ◽  
Scale Test ◽  
Screening Sensitivity

This paper describes the uncertainty analysis carried out for 10% Hot leg break LOCA of Large Scale Test Facility as a part of IAEA Coordinated Research Project on “Evaluation of Uncertainty in Best Estimate Accident Analysis”. The best estimate code used for this analysis is RELAP5/MOD3.2. Initially the nodalisation of the test facility for carrying out the analysis is qualified for both steady state and transient level by systematically applying the procedures lead by Uncertainty Methodology based on Accuracy Extrapolation developed at University of Pisa. Subsequently the uncertainty analysis is carried out using sampling based Monte Carlo approach, which involves the generation and extrapolation of a mapping from uncertain inputs to the uncertain analysis results. The major steps followed in this methodology mainly includes screening sensitivity analysis for input parameters, design matrix generation using Latin Hypercube Sampling, representation of uncertainty analysis results based on best estimate thermal hydraulic code runs and importance /sensitivity analysis using regression analysis. The steps followed have been described in details in this paper.

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 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.

The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

2002 ◽  
Author(s):  
Salih Gu¨ntay ◽  
Abdel Dehbi ◽  
Detlef Suckow ◽  
Jon Birchley
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Severe Accident ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Full Size ◽  
Management Actions ◽  
Accident Management ◽  
Dry Conditions ◽  
Secondary Side

Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes’ models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an unisolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (AeRosol Trapping In a Steam generaTor), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2) the near field close to break under dry conditions, 3) the bundle far-field under dry conditions, 4) the separator and dryer under dry conditions, 5) the bundle section under wet conditions, 6) droplet retention in the separator and dryer sections and 7) the overall SG (integral tests). Prototypical test parameters are selected to cover the range of conditions expected in severe accident as well as DBA scenarios. This paper summarizes the relevant issues and introduces the ARTIST facility and the provisional test program which will run between 2003 and 2007.

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