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.     

Serviceability design load factors and reliability assessments for reinforced concrete containment structures

A reinforced concrete nuclear power plant containment structure is subjected to various random static and stochastic loads during its lifetime. Since these loads involve inherent randomness and other uncertainties, an appropriate probabilistic model for each load must be established in order to perform reliability analysis. The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops probability-based load factors for the limit state design of reinforced concrete containment structures. The purpose of constructing reinforced concrete containment structure is to protect against radioactive release, and so the use of a serviceability limit state against crack failure that can cause the emission of radioactive materials is suggested as a critical limit state for reinforced concrete containment structures. Load factors for the design of reinforced concrete containment structures are proposed and carried out the reliability assessments.     

Assessment of SAFER03 Code Using ROSA-III Break Area Spectrum Tests on Boiling Water Reactor Loss-of-Coolant Accidents

The SAFER03 computer code has a newly developed evaluation model for the analysis of various boiling water reactor (BWR) loss-of-coolant accidents (LOCAs). Analyses of the ROSA-III break area spectrum tests in a recirculation line were performed using the SAFER03 to assess the predictive capability of the code for a BWR LOCA. The ROSA-III test facility at the Japan Atomic Energy Research Institute (JAERI) was constructed to simulate a LOCA in a BWR/6-251 plant with 848 fuel bundles and 24 jet pumps. This paper summarizes the assessment results of SAFER03 which predicted the system responses and key phenomena well and the conservative peak cladding temperature (PCT) for recirculation line break tests with different break areas.     

Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to ≈40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work.     

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.     

The vent-to-vent desynchronization effects on LOCA steam condensation loads in BWR pressure suppression pool

During a loss-of-coolant accident of a boiling water reactor (BWR), direct-contact steam condensation in the pressure suppression pool will induce undesirable dynamic pressure loads onto pool boundary structures. The magnitudes of the pool boundary loads will be influenced by finite desynchronization among the condensation processes, which take place at the outlets of multiple (about one hundred) vent pipes. This paper investigates the effects of the condensation desynchronization on the pool boundary loads, with regard to test data obtained from large-scale tests which represented seven full-sized vent pipes in a BWR Mark II pressure suppression containment system. The desynchronization effects are studied by analyzing the test data in the time and frequency domains. Attempts are made to extrapolate the experimental results to the actual plant geometry. Mechanisms responsible for the desynchronization are also discussed.     

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.     

Simple Method of Eliminating Infinite Variance in Point Detector Problem of Monte Carlo Calculation

Containment spray tests for elemental iodine vapor removal were conducted under simulated BWR LOCA condition with large-scale JAERI Model Containment. The tests were carried out to determine the behavior of gas phase iodine removal by spray water containing carbonic acid, the mixing effect in the sump pool, and iodine deposition onto the containment wall.

In the tests, a half-life of the iodine concentration decrease in gas phase was obtained as 6–12min and an iodine partition coefficient was also obtained as 80–160 by the spray washout under pH range of 5.3–5.9. The effect on the mitigation of pH decrease of spray water caused by the iron corrosion process was discussed from the experiments. Results obtained from the tests were compared with those calculated by a code MIRA-PB.     

Exact Solutions of Two Point Time Correlations on Point Reactor Kinetics with Gaussian Reactivity Fluctuations

Station blackout is reported to be a sequence that would likely be a significant contributor to the accident risk at a boiling water reactor (BWR). The occurrence frequency of station blackout is evaluated in probabilistic safety assessment (PSA) to be 6×10^?6 per reactor year at Limerick and less than 10^?7 per reactor year at BWR in Japan.

This report describes an analytical study of thermal-hydraulic and radionuclide behavior during a postulated severe accident of station blackout at a reference BWR plant. The analytical approach was shown in both of hand calculation and the THALES/ART code calculation to better understand wide physical and chemical phenomena in the processes of severe accidents.

We evaluated timing of key events, core cooling and core temperature, reactor vessel failure, debris temperature, containment pressure, and release and deposition of radionuclide in the containment. The THALES and CORCON models on the chemical reactions in the core-concrete interaction lead to great differences in the increasing rate of containment pressure and the release rate of fission products from the core debris.     

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.     

Development of MINCS Code for Evaluation of Two-Phase Flow Models†

Abstract

This report describes the development of the one-dimensional two-phase flow characteristics analysis code, MINCS. The MINCS has been developed at JAERI (Japan Atomic Energy Research Institute) to provide a computational tool for analyzing not only physical phenomena of two-phase flows but also numerical phenomena in two-phase flow analyses. In MINCS, nine types of two-phase flow models, ranging from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model, can be treated under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected or supplied by users. It is, therefore, easy to solve physical or numerical benchmark problems.     

Analyses of Pool Swell Tests by Two-Dimensional Hydrodynamic Computer Code

A two-dimensional hydrodynamic computer code SOLA-VOF was examined on the analytical capability for dynamic loads by pool swell in the MARK-I type BWR suppression chamber under LBLOCA (Large Break Loss of Coolant Accident) conditions. Two pool swell tests, (LLL 1/5-scale and EPRI 1/12-scale tests) were selected for this purpose and analyzed by the SOLA-VOF code modified with incorporation of a simple downcomer flow model. In these analyses, it was necessary to take account of three-dimensional effect of pool swell behavior along the chamber axis by use of a method such as spatially weighting function experimentally determined, because a simple two-dimensional calculation by the SOLA-VOF code gave too much conservative evaluation for the impact load on the ring header. Applications of this method gave a good agreement between the calculation and measurement. The vertical loads on the suppression chamber wall were well analyzed by this code. It might be because the local pressure difference caused by the nonuniform pool swelling disappeared owing to pressure integration on the surface of suppression chamber wall.     

Calculation of Iodine Removal by Spray in LWRs Containment Vessels

A computer code MIRA-PB for predicting the iodine removal by containment spray in LOCA was prepared on the basis of MIRA-P/MIRA-B code developed in Battelle Columbus Laboratories. MIRA-PB considers behavior of inorganic iodine, organic iodide, and iodic aerosol and simultaneous removal by natural deposition, liquid-film absorption, spray washout, filtration and leakage to the environment. The iodine removal by the containment spray systems in LOCA of PWR and BWR is calculated with the MIRA-PB.     

Analytically predicted versus measured response of a free-standing steel containment vessel subjected to safety-relief valve discharge loads

Following the actuation of safety-relief valves in BWR nuclear power plants, first water then air and steam are cleared from the discharge lines through quencher devices into a suppression pool. This clearing results in water spike, air bubble, and condensation pressure loads applied to structures in the pool, and the surrounding containment vessel.The Leibstadt Nuclear Power Plant has the only free-standing steel Mark III containment vessel in the world. All other steel Mark III containment vessels have concrete backing in the suppression pool region, which dampens clearing load responses. As such, it is of interest to note how this steel vessel responds to discharge pressures, and compare these responses to analytically predicted results.The purpose of this paper is to compare the analytical results used to design the steel containment vessel with the responses measured during in-plant testing. The analytical methods considered the effects of fluid-structure interaction. The test program included initial and consecutive actuations of a single valve, and initial actuation of multiple (four) valves. The conclusion of the comparison is that, in general, there are large conservatisms in the analytical predictions versus measured responses.     

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

Experimental verification of accident loads for PWR containments

The principal loads which a nuclear power reactor containment is designed to withstand are produced by internal fluid caused static and/or dynamic pressures. They can be generated during failure events which release mass and energy into the containment atmosphere. An overview of the events which can generate substantial internal loads is provided. Representative experimental programs initiated for the investigation of the relevant physical phenomena are described. Illustrative examples of measured data are presented and discussed.     

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10^?2 (mean) and 3.9x 10^?2 (median) for the BWR suppression pool case, 2.2x10^?3 (mean) and 2.8x10^?10 (median) for the BWR pedestal case, and 6.8X10^?2 (mean) and 1.4x10^?2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.     

Heat Removal Performance Evaluation of Several Passive Containment Cooling Systems during Loss of Coolant Accident

As a passive containment cooling system (PCCS), which is adopted in simplified BWRs, several concepts, differing in cooling location and method, such as the suppression chamber water wall, the drywell water wall, the isolation condenser (I/C) and the drywell cooler, have been considered. This paper summarizes the characteristics of each PCCS concept, and the analysis results of the performance for several PCCSs during a main steam line break LOCA for a reference simplified BWR plant, obtained by the newly developed containment thermalhydraulic response analysis code TOSPAC.

The performance comparison suggests that I/C and drywell cooler have good heat removal capability with regard to the smallest heat transfer area among PCCS concepts evaluated in the present analysis. I/C removes decay heat efficiently, since it absorbs steam directly from the reactor pressure vessel, which is the hottest portion inside the containment. The suppression chamber water wall is ineffective, mainly due to high non-condensable gas partial pressure in the suppression chamber, and low suppression pool temperature.

Calculations of other pipe breaks were also implemented for the reference plant adopting I/C as PCCS. The results show the effectiveness of the I/C cooling over a wide range of break spectra.