Overview of SAFER03 Assessment Studies

The safety research for BWRs has been positively done by the JAERI, Japanese BWR utilities and BWR vendors in this decade and has shown the important phenomena under BWR LOCA conditions. Based on these significant results, the SAFER03 computer code was jointly developed by Toshiba, Hitachi and General Electric. SAFER03 has been qualified against the BWR simulation test data obtained from TBL, ROSA-III and FIST-ABWR test facilities. The objectives of this study are to assess the predictive capability of SAFER03 code to simulate the significant LOCA phenomena and to catch key parameters during BWR LOCA. This paper summarized the results of these SAFER03 assessments and showed that SAFER03 could predict the realistic behavior of BWR LOCA with slight conservative peak cladding temperatures.     

Similarity study of ROSA-III and fist large break cuonterpart tests to BWR large break LOCA

A large break test in a recirculation pump suction line with the assumption of LPCI-diesel generator failure was conducted at the ROSA-III test facility of Japan Atomic Energy Research Institute. A counterpart test was also performed at the FIST test facility of General Electric Company. The objective of the tests was to develop common understanding and interpretation of the controlling thermal-hydraulic phenomena during a large break LOCA of a BWR. The fundamental thermal-hydraulic phenomena in the ROSA-III and FIST tests such as the system pressure, mixture level and fuel rod surface temperatures agreed well. The FIST test had more bundle uncovery than that in ROSA-III since lower plenum steam in the FIST test flowed out of the jet pumps when they uncovered allowing more liquid to drain from the bundle. The ROSA-III and FIST tests and a BWR counterpart were analyzed with the RELAP5/MODI (cycle 018) code. The similarity of the ROSA-III and FIST large break tests to a BWR large break LOCA has been confirmed through comparison of calculated results though they are slightly different in details. It is perhaps desirable to reexamine the DNB and interphase drag correlations and the jet pump models usedin the code.     

BWR small break LOCA counterpart tests at ROSA-III and FIST test facilities

The Japan Atomic Energy Research Institute performed a 2.8% recirculation pump suction line break BWR LOCA test at the ROSA-III test facility. The test was a counterpart test to the 2.8% break test performed at the FIST test facility by the General Electric Company. The objective of the test was to develop a common understanding and interpretation of the controlling phenomena for a small break LOCA of a BWR. Similar phenomena were observed in the two tests in a similar time sequence and with magnitudes. These two test results and a 2.8% break reference BWR LOCA were analyzed using the THYDE-B1 computer code. It was confirmed from the analysis that the THYDE-B1 code has enough capability to analyze a BWR small break LOCA. The applicability of the tests performed at the two facilities to a BWR was also confirmed through the analyses.     

Analysis of large break LOCA experiment at ROSA-III

Analysis of the ROSA-III test RUN 704 was performed by using the computer codes RELAP4J, RELAP4/MOD6 and RELAP5/MOD0 to verify the predictive capability of the codes for a BWR LOCA. The ROSA-III facility is a volumetrically scaled (1/424) BWR system with an electrically heated core, designed for in tegral LOCA/ECCS tests. The RUN 704 experiment at the ROSA-III test facility simulated a 200% double-ended offset shear break on the inlet side of the pump in the recirculation loop. From present analyses, key parameters which are important to predict major behavior during a BWR large break LOCA have been clarified and the promising predictive capability of the advanced code RELAP5 has been verified.     

Investigation of BWR LOCA at ROSA-III — Effect of break configuration on system transients

The ROSA-III test facility is a volumetrically scaled ( ) BWR/6 system with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA).Six loss-of-coolant experiments with a break area of 15%, 50% or 200% at the main recirculation pump inlet line were conducted at the ROSA-III test facility with a high pressure core spray failure. A sharp-edged orifice or a long throat nozzle was used as a break plane. It was found in the experiments that the break flow differences between the orifice and the nozzle break configurations with the same flow area were observed only in the subcooled break flow region. Subcooled break flow rate through the orifice was much larger than that through the nozzle. The break configuration difference had little influence on the other system responses, especially on the peak cladding temperature. The applicability of the test results to a BWR/6 has been confirmed through analyses of the 15% break ROSA-III LOCA experiments and BWR/6 LOCAs by using RELAP4/MOD6/U4/J3 code. The experimental results of the ROSA-III LOCA experiments were calculated well by the code, and the same trends were calculated in the BWR analyses.     

SAFER03 and TRAC-BD1 analyses of TBL experiments on a boiling water reactor loss-of-coolant accident

Simulation tests of a BWR LOCA were conducted with a Two Bundle Loop (TBL). Japanese BWR utilities, Toshiba Corp. and Hitachi Ltd. have jointly performed the SAFER03 and TRAC-BD1 code analyses with the TBL test data. The objective of this study is to demonstrate the predictive capability of SAFER03 and TRAC-BD1 for a BWR LOCA. This paper summarizes the results of the analysis in which SAFER03 and TRAC-BD1 predicted the system responses and key phenomena well and a conservative peak cladding temperature of SAFER03 was obtained.     

Boiling Water Reactor Loss of Coolant Tests Single Failure Tests with ROSA-III

The single failure tests with the ROSA-III were simulated BWR LOCA experiments by the scaled BWR test facility resulting from a 200% double-ended break at the recirculation pump suction line to evaluate the core cooling capability of a BWR ECCS under the single failure condition.

The experimental results showed that the loss of LPCS and one LPCI resulted in the highest PCT of 870 K of the single failure series tests, yet a core cooling capability by the ECCS was maintained. The REALP4/Mod 6 code was used to evaluate the predictive capability of the LOCA analysis code. The calculated results showed that the RELAP4/Mod 6 code was able to predict occurrences and sequence of major events anticipated to occur during a BWR LOCA correctly. However it was found that the code still needs to be improved in a CCFL model to better describe thermohydraulic behavior in the core.

The analyses presented in this paper are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a BWR during a LOCA.     

Five percent break BWR LOCA/ECC test at ROSA-III without HPCS actuation - two-dimensional core thermal-hydraulic phenomena

The ROSA (Rig of Safety Assessment)-III facility is a volumetrically scaled (1/424) simulated boiling water nuclear reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. A recirculation pump suction line break test with a five percent break area was conducted with the assumption of high pressure core spray system (HPCS) failure. The simulated peripheral fuel rods facing the channel box wall had a tendency to be rewetted temporarily at the upper part of the core by falling water from the upper plenum before low pressure core spray system (LPCS) actuation, while the rods in the central region were not rewetted but quenched mainly from the bottom of the core after low pressure coolant injection system (LPCI) actuation. Therefore, the peak cladding temperatures of the simulated high power fuel rods were limited to lower values since they were located in the peripheral region and the temporary rewetting before LPCS actuation occurred mainly in the peripheral region. The ROSA-III five percent break test and a BWR counterpart were analyzed with the RELAP5/MOD1 (cycle 018) code. Similarity between the ROSA-III small break test and a BWR small break LOCA has been confirmed through comparison of the calculated results.     

Similarity study of large steam line break LOCA's in ROSA-III, FIST and BWR/6

Similarity of the thermal hydraulic phenomena in a 100% steam line break loss-of-coolant accident (LOCA) between the Rig-of-Safety Assessment (ROSA)-III. Full-Integral Simulation Test (FIST) and a boiling water reactor (BWR)/6 system has been studied experimentally and analytically. The experimental results of ROSA-III (RUN952) and FIST (6MSB1) showed similar LOCA phenomena except for the core cooling. The core cooling was affected by the different ECCS actuation logics used in the tests. The effects of the different test conditions and the system-inherent features on the LOCA phenomena were separately evaluated through the post-test and similarity analysis of the ROSA-III and FIST tests by using RELAP5/MOD1 code with a jet pump model. The similarity of the major events in the ROSA-III and FIST facilities to those of BWR/6 system were confirmed assuming the same ECCS actuation logic and the same sealed initial mass inventory among the three systems. Differences in vessel geometries, metal stored heat and core power curves caused slight differences in the responses of pressure and fuel surface temperatures.     

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

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. 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.     

Investigation of Break Location Effects on Thermal-Hydraulics during Intermediate Break Loss-of-Coolant Accident Experiments at ROSA-III

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.     

Simulation tests of BWR non-loca transients using the tbl

In the last decade, a large number of experiments have been performed in order to understand the thermal-hydraulic response in a boiling water reactor (BWR) under postulated loss of coolant accident (LOCA) conditions. These experimental results showed that the core cooling effect under the LOCA conditions was significantly affected by three-dimensional and multi-bundle phenomena after emergency core cooling systems (ECCSs) started. Also, the peak cladding temperature (PCT) during the LOCA was kept below a specific value of the licensing acceptance criteria, 1473 K (1200°C). These key results of the experiments were incorporated into a computer code, SAFER, which was developed for the BWR LOCA/ECCS analyses under the cooperative studies of Hitachi Ltd, Toshiba Co., and General Electric Co. (GE).

In a couple of years, the experimental study of multi-bundle phenomena was extended into the area of off-normal and non-LOCA transients. Thermal-hydraulic responses during boiling transition were studied using the TBL (Two Bundle Loop) test facility with two full-length bundles. The experimental results showed that interaction and feedback effects between the bundles were expected to be unaffected by core cooling during the typical off-normal and non-LOCA transients. Also, the SAFER showed good predictions for hydraulic responses in the bundles and temperature transients of the rod surfaces.     

Best Estimate BWR Transient Analysis with TRACG Assessment using Data from BWR Startup Tests and LOCA Integral Tests

TRACG is a new version of the best estimate BWR transient analysis code, which utilizes a multi-dimensional two-fluid model for the thermal hydraulics and a three-dimensional neutron kinetics model. A three-dimensional neutronics, a fully implicit integration scheme and models for advanced BWR components have been implemented in the code upon TRAC-BF1.

Assessment of TRACG has been performed in this study for the predictive capability of plant transients, which include thermal-hydraulic and neutronic interactions, as affected by responses of the plant control system. Simulations were presented for BWR representative transient tests, which were done as part of a series of BWR5 startup tests. As for the capability to predict thermal hydraulics during the design basis LOCAs, simulations were presented for the LOCA integral tests conducted in the ROSA-III at JAERI and the Hitachi TBL, which had been used for assessment of the TRAC former version.

Consequently, (1)the space-dependent power flow transitions in a BWR were confirmed by TRACG simulations in which the module coupled with neutronics and thermal hydraulics during transients has been newly introduced, and (2) the characteristic thermal-hydraulic phenomena including multi-channel effects during the design basis LOCAs were confirmed, as well as the TRAC former version, by TRACG simulations on which the influence due to a fully implicit integration scheme has not extended. Capability of TRACG to predict BWR transients ranging from simple plant operational transients to design basis LOCAs was successfully demonstrated.     

Spray Cooling Test

A steady separate effects test on BWR spray cooling was performed at relatively high system pressures using the ROSA-III test vessel. These tests were conducted in order to promote a better understanding of the thermal-hydraulic phenomena in LOCA experiments and to obtain information necessary for improvement of analytical codes. The fraction of entrainment or overflow for various spray conditions was obtained and the data of CCFL at the upper tie-plate were compared with correlations. It was shown that the occurrence of CCFL significantly diminished core cooling effects and that rod quench by fall back water was quite irregular and unstable. Reflood core cooling was also studied.     

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

The ROSA-111 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/m²K. It is larger than existing data measured at lower pressures during a spray cooling period. Bottom-up quench temperatures are given by a simple equation: the sum of the saturation temperature and a constant of 262 K. Then the heat transfer model in the RELAP4/MODE/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.     

Investigation of fluid mixing with direct vessel injection

The object of this work is to investigate fluid mixing phenomena as they related to pressurized thermal shock (PTS) in a pressurized water reactor vessel downcomer during transient cooldown with direct vessel injection (DVI), using test models. The test model designs were based on ABB Combustion Engineering (CE) System 80+ reactor geometry. A cold-leg, small-break loss-of-coolant accident (LOCA) and a main steam line break were selected as the potential PTS events for the ABB-CE System 80+. This work consists of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluids and existing coolant in the downcomer region, and the second part presents the results of thermal mixing tests with DVI in the other test model. Flow visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small-break LOCA. The measured transient temperature profiles compare well with the predictions from the REMIX code for a small-break LOCA, and with the calculations from the COMMIX-1B code for a stream line break event.     

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

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

Iodine Removal Tests for BWR Containment Spray by Large Scale Facility

Iodine removal tests for a BWR containment spray were carried out with large-scale JAERI Model Containment Test Facility under LOCA simulated conditions. The tests consisted of two groups: “gas-phase based” tests mainly to obtain the initial iodine removal rate by the spray and “liquid-phase based” tests to obtain the iodine partition coefficient at equilibrium state. It was shown that the degree of iodine removal was largely influenced by pH-value of spray water. The results were discussed with calculated results by a code MIRA-PB using a dose reduction factor for the airborne iodine.     

Blowdown thrust force under pipe rupture accident: II. Analytical evaluations of partial break and ramped opening break

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.     

Blowdown thrust force under pipe rupture accident: I. Experimental evaluations of blowdown thrust force and decompression characteristics

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.