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.

Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor

2014 ◽  
Author(s):  
Shinya Miyata ◽  
Satoru Kamohara ◽  
Wataru Sakuma ◽  
Hiroaki Nishi
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Core Cooling

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.

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.

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.

CFD Analysis of Subcooled Wall Boiling at Shell Side of Steam Generator With TSP

2014 ◽  
Author(s):  
Chenglong Wang ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Wenxi Tian ◽  
Guanghui Su
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Steam Generator ◽  
Flow Boiling ◽  
Nucleate Boiling ◽  
Two Phase ◽  
Pressurized Water ◽  
Support Plate ◽  
Coupled Heat Transfer ◽  
Secondary Side

This paper addresses the numerical simulation of two-phase flow heat transfer among the tube bundles with tube support plate (TSP) of an integral type pressurized water reactor steam generator using RPI wall boiling model. The subcooled nucleate boiling phenomenon and the coupled heat transfer between the SG primary side and secondary side were obtained. Also, the effects of tube support plate (TSP) and the different inlet subcooling on the thermal-hydraulic characteristics of SG were studied. From the results of the present numerical simulation, it reasonably revealed the subcooled flow boiling occurred in the SG secondary side and the distributions of key parameters around TSP, elucidating that this model can provide useful information to the design of the steam generator.

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.

RBHT Steam Cooling and Droplet Injection Pre-Test Analysis Using COBRA-TF

2001 ◽  
Author(s):  
C. Frepoli ◽  
A. J. Ireland ◽  
L. E. Hochreiter ◽  
F. B. Cheung
Keyword(s):  
Heat Transfer ◽  
Local Heat Transfer ◽  
Injection System ◽  
Test Facility ◽  
Film Boiling ◽  
Two Phase ◽  
Field Representation ◽  
Dispersed Flow ◽  
Spacer Grids ◽  
Flow Film Boiling

Abstract The droplet injection experiments to be performed in a 7 × 7 rod bundle heat transfer test facility are being simulated using an advanced thermal hydraulics computer code called COBRA-TF. A current version of the code, which provides a three-dimensional, two-fluid, three-field representation of the two-phase flow, is modified to facilitate the simulation of the droplet field produced by the injection system in the test facility. The liquid phase is split into a continuous liquid field and droplet field where a separate momentum and mass equation is solved for each field, with the effects of spacer grids being properly accounted for. Pre-test analyses using the modified COBRA-TF code have been conducted for different injection conditions. Results indicate that there are specific ranges of conditions that can be simulated within the facility constraints to provide for validation of the dispersed flow film boiling models. The numerical results also show important effects of the spacer grids on the local heat transfer in the dispersed flow film boiling regime.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break1

2015 ◽  
Vol 137 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Pipe ◽  
Short Period ◽  
Main Steam Line ◽  
Main Steam ◽  
Transient Thermal ◽  
Secondary Side

This paper presents a multidimensional numerical analysis of the transient thermal-hydraulic response of a steam generator (SG) secondary side to a double-ended guillotine break of the main steam line attached to the SG at a pressurized water reactor (PWR) plant. A simplified analysis model is designed to include both the SG upper space, which the steam occupies and a part of the main steam line between the SG outlet nozzle and the pipe break location upstream of the main steam isolation valve. The transient steam flow through the analysis model is simulated using the shear stress transport (SST) turbulence model. The steam is treated as a real gas. To model the steam generation by heat transfer from the primary coolant to the secondary side coolant for a short period during the blow down process following the main steam line break (MSLB) accident, a constant amount of steam is assumed to be generated from the bottom of the SG upper space part. Using the numerical approach mentioned above, calculations have been performed for the analysis model having the same physical dimensions of the main steam line pipe and initial operational conditions as those for an actual operating plant. The calculation results have been discussed in detail to investigate their physical meanings and validity. The results demonstrate that the present computational fluid dynamics (CFD) model is applicable for simulating the transient thermal-hydraulic responses in the event of the MSLB accident including the blowdown-induced dynamic pressure disturbance in the SG. In addition, it has been found that the dynamic hydraulic loads acting on the SG tubes can be increased by 2–8 times those loads during the normal reactor operation. This implies the need to re-assess the potential for single or multiple SG tube ruptures due to fluidelastic instability for ensuring the reactor safety.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

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