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.     

Development of Evaluation Code for Long-Term Cooling of Containment with Passive Safety Feature

Simplified BWRs are characterized as an adoption of a passive ECCS and a passive containment cooling system (PCCS). While a passive ECCS has a short term core cooling function, a PCCS has a long-term decay heat removal function. As a PCCS, several concepts, differing in cooling location and method employed, have been considered. From the containment thermal- hydraulic response analysis viewpoint, simplified BWRs are essentially different from the current BWRs. For evaluating and comparing the performance of several PCCSs over full break spectra, the new containment safety evaluation code TOSPAC was developed as a preliminary design tool for PCCS. This paper summarizes the thermal-hydraulic modelings of the TOSPAC code and the validity evaluation of the TOSPAC code, compared with TRAC-BF1 calculation.

From the validity evaluation concerning a main steam line break (MSLB) accident analysis for an isolation condenser (I/C) as a PCCS, it was found that the TOSPAC calculation result shows reasonable agreement with that for TRAC, even though the TOSPAC consists of simpler modelings.     

Upper Plenum Dump during Reflood in PWR Loss-of-Coolant Accident

Upper plenum dump during reflood in a large break loss-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood.

The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnitude of water subcooling.     

Restrictive Effect of Ascending Steam on Falling Water during Top Spray Emergency Core Cooling

Water spraying experiments were conducted to find out a flow rate of falling water overcoming ascending steam during top spray emergency cooling with an 8×8 type simulated fuel rod bundle of real size. The bundle consisted of 64 rods, each with a diameter of 12.5 mm, arranged in the form of square lattice with a pitch of 16.3 mm. In the experiments the simulated fuel rods were not heated. Instead, steam was injected into the lower plenum vessel simulating bundle-generated steam. As the results, (1) a criterion was proposed to determine the region where the restrictive effect of ascending steam on falling water appears, considering the decrease of a flow rate of ascending steam due to condensation by a spray of subcooled water, (2) the restrictive effect was independent of water head on the upper tie plate and water injection methods, and (3) an analytical model based on the pressure balance at the upper tie plate was proposed to calculate a flow rate of falling water overcoming ascending steam.     

Multi-Channel Thermal Hydraulic Model for LOCA Analysis of Heterogeneous BWR Core

A multi-channel thermal hydraulic model for LOCA analysis of a heterogeneous core such as a HCBWR has been developed. This model solves integral formulations for basic equations based on a one-dimensional drift flux model. The core region is divided into several fuel channel groups which differ in their thermal power or geometry. The various flow patterns in the core are determined by calculating the redistribution of vapor generated in the lower plenum into the fuel channel groups. In order to verify the multi-channel model, a computer program FLORA was developed based on the multi-channel model and large and small break LOCA experiments conducted in the Two Bundle Loop (TBL) facility were analyzed by the FLORA program. As a result, the difference in thermal hydraulic behavior between two bundles with different power in the various break LOCA experiments were well simulated.     

Effect of Loop Seal on Reflood Phenomena in PWR

The blockage of the primary coolant loop with water filled at the crossover legs, denoted “loop seal”, is expected to give a significant influence on core cooling during the reflood phase of a PWR-LOCA. However, the effect of the loop seal has been little investigated. Therefore, it was studied experimentally by using Cylindrical Core Test Facility (CCTF).

The loop seal was cleared in a short time (40–60 s) after reflood initiation by pushing-off of stagnating water in the crossover legs due to steam accumulation and resultant pressure increase in the upper plenum.

Although the core cooling was degraded during the loop seal period, it recovered after the loop seal clearing. The degradation of core cooling during the loop seal period is considered to be caused by the low core-inlet water flow rate and resultantly by the small volumetric fraction of water in the core.

The quantitative estimation about the loop seal effect on the clad temperature was made and it was indicated that the maximum clad temperature would not exceed the allowable upper limit (1,473K) specified in the licensing about the reactor safety even with the loop seal at the beginning of the reflood phase.     

Visual Study of Flow Behavior in Upper Plenum during Simulated Reflood Phase of PWR-LOCA

Fuel temperature is an important parameter in reactor safety. However, temperatures of fuel sub-assemblies in power reactors usually cannot be directly measured. A means therefore has been devised to use heat removed by coolant from fuel sub-assemblies following a reactor trip to estimate the pre-trip average fuel temperatures and fuel-to-coolant heat transfer coefficients. Sub-assembly coolant mass flow and outlet temperature measurements are used to determine the heat removed. After accounting for the contribution from release of stored heat from non-fuel components and from decay heat, the contribution due to release of fuel-stored heat alone during the trip transient can be inferred. Average fuel temperatures and fuel-to-coolant heat transfer coefficients of the FBR “MONJU” at 45% thermal power were estimated using this energy balance approach. The trip-test data derived estimates compared favorably with plant simulation code calculated values.     

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.     

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.     

Two-Phase Flow Phenomena in Broken Recirculation Line of BWR

When a primary recirculation line of BWR is ruptured, a primary recirculation pump may be subjected to very high velocity two-phase flow and its speed may be accelerated by this flow. It is important for safety evaluation to estimate the pump behavior during blowdown. There are two problems involved in analyzing this behavior. One problem concerns the pump characteristics under two-phase flow. The other involves the two-phase conditions at the pump inlet. If the rupture occurs at a suction side of the pump, choking is considered to occur at a broken jet pump nozzle. Then, a void fraction becomes larger downstream from the jet pump nozzle and volumetric flow through the pump will be very high. However, there is little experimental data available on two-phase flow downstream from a choking plane. Blowdown tests were performed using a simulated broken recirculation line and measured data were analyzed by TRAC-PIA. Analytical results agreed with measured data.     

RELAP4 Modeling for Reverse Flow Property of BWR Jet Pump under Large Break LOCA Condition

It is very important to identify the reverse loss coefficient of BWR jet pump in the evaluation of core inlet flow at the beginning phase of BWR LBLOCA (Large Break Loss-of-Coolant Accident) analyses. Hence, the reverse flow property of jet pump was investigated in relation between the momentum equation, pressure loss coefficient and RELAP4 noding, and a new modeling has been proposed. In the proposed modeling, an equivalent pressure loss coefficient is used to take into account of the effect of accellerating pressure loss by the continuous flow area reduction from the tale pipe to the throat. The effectiveness of this model was studied by analyses for the LOFT 1/6 scale jet pump experiment and typical BWR LBLOCA. It has been, consequently, shown that this proposed model gives better jet pump property than a previous model which is used in the WREM sample problem and which gives very conservative result in core inlet flow and in the peak cladding temperature through whole transient.     

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.     

LOCA下喷放参数对冷却剂喷放特性影响数值研究

冷却剂喷放过程是失水事故（LOCA）的重要过程之一,研究冷却剂喷放过程的热工水力特性对认识LOCA以及预测事故后放射性源项迁移过程有着重要意义。本文利用FLUNET软件建立冷却剂喷放数值计算模型,并对其进行验证。利用模型研究喷口直径、喷放距离和喷放压力等喷放参数对计算域内流场温度、液滴速度和蒸汽流速等特性的影响。研究结果表明：喷口直径的提高使得喷放参数均有提高;随喷放距离的增大,流场温度和液滴速度先上升后下降,而蒸汽流速先上升后趋于平稳;喷放压力越大,喷放参数的最大值离喷放出口越远,液滴速度和蒸汽流速的最大值随喷放压力的增大逐渐上升,而流场温度最大值没有变化。     

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.     

Effects of ECC Water Injection Rate on Reflood Phase during LOCA

Some metal iodides such as of Fe, Al, Zr and Te are known to cause stress corrosion cracking (SCC) of Zircaloy just as iodine itself does. Therefore 15 metal iodides were selected as corrodants, and SCC tests were carried out using the internal gas pressurization method.

The results showed that: (1) only those metal iodides which react thermodynamically with Zr to produce ZrI₄ cause SCC of Zircaloy-2; (2) when SCC occurs, the reaction rate between the iodide and Zr seems to be a main factor in determining the SCC susceptibility; (3) gaseous ZrI₄ is the most corrosive agent; and (4) some species of metal iodides, such as PbI, cause SCC of Zircaloy-2 more easily than I₂ vapor.

Scanning electron microscope (SEM) examination and electron probe microanalysis (EPMA) on the fracture surface of failed specimens revealed that ZrI₄, formed as the reaction product between the metal iodides and Zr, might induce SCC of Zircaloy-2 rather than the iodides themselves.     

Effect of Higher Order Anisotropy on Disadvantage Factor in a Cell

Among various fluctuating quantities in a BWR plant the present authors dealt with core channel flow fluctuations. Using the instrumented fuel assemblies (IFA) installed in the Japan Power Demonstration Reactor II core, fluctuations of the inlet and outlet channel flow rates were observed at power operation condition. Correlation analyses of the fluctuating signals revealed the followings: (a) the flow fluctuations in any IFA channel showed almost uncorrelated cross-correlation with other IFA channel flow, (b) cross-correlation between the neutron density fluctuations and each IFA channel flow fluctuations also showed little correlation, and (c) the forced circulation pump flow showed unexpectedly weak correlation with the core channel flow.

To attempt to explain the above facts, some hypotheses are introduced: (1) Each channel will be actuated by an independent weak noise source. (2) The fluctuating flow of a channel will interfere with each other, then the interferences will spread to the whole core and balance in a stationary state forming a certain statistical pattern. A dynamic equation of the channel flow fluctuations is introduced based on the above hypotheses, and the covariance functions of inter-channel flow fluctuations are calculated. The hypotheses are discussed by comparing with the results of identification by the use of an autoregressive model.     

Numerical Study on Influence of Blowdown Parameter on Coolant Blowdown Characteristic in LOCA

The coolant blowdown process is one of the important processes of the loss of coolant accident (LOCA). It is of great significance to study the thermal hydraulic characteristics of coolant blowdown process for understanding LOCA and predicting the migration process of radioactive source term after accident. The numerical simulation model of coolant blowdown was established by FLUENT software and verified. The model was used to study the effects of blowdown parameters such as diameter of nozzle, blowdown distance and blowdown pressure on flow field temperature, droplet velocity and vapor velocity. The results show that the increase of diameter of nozzle increases the blowdown parameters. As the blowdown distance increases, the flow field temperature and the droplet velocity increase first and then decrease, while the vapor velocity first rises and then stabilizes. The greater the blowdown pressure is, the farther the blowdown parameter is from the blowdown outlet. The maximum values of droplet velocity and vapor velocity increase gradually with the blowdown pressure, while the maximum value of the flow field temperature does not change.     

Experimental Study on Heat Removal Characteristics for Water Wall Type Passive Containment Cooling System

To evaluate the heat removal capability of a water wall type cooling system, which is one passive containment cooling system (PCCS), the thermal hydraulic behavior in the suppression pool (S/P) and the outer pool (O/P, flat plate water wall) have been investigated experimentally. The following results were obtained. (1) A thermal stratification boundary, which separates the pools into the upper high temperature and lower low temperature regions, was formed just below the vent tube outlet. (2) Convection heat transfer characteristics in the S/P and O/P along the primary containment vessel (PCV) wall had no significant differences and were those of natural convection. Correlation of the natural convection heat transfer up to the Ra number of 2×10¹⁴ was obtained. (3) Vertical variations of local condensation heat transfer coefficients under a noncondensable gas presence were within ±10% of the average value for the 4.7 m heat transfer length. An experimental correlation for the average condensation heat transfer coefficients was obtained as a function of steam and noncondensable gas mass ratio. (4) An analytical model to evaluate the system performance of the water wall type PCCS was verified. (5) A baffle plate concept to mitigate thermal stratification at the vent outlet and to enlarge the high temperature region in the S/P was considered as a means to improve heat release capability. Thermal hydraulics with a baffle plate were examined, and effectiveness of the baffle plate to improve the heat release capability was confirmed.