RETRAN's improved small break modeling capability

The RETRAN-02 computer code has been used extensively to analyze both operational transients and small break loss-of-coolant accidents (SBLOCA). While the results of these analyses are generally in good agreement with data, the explicit solution method sometimes requires small time-steps for portions of the small break calculations. The RETRAN-03 code, now under development, is incorporating additional models to extend the capability of RETRAN-02 and improve the code efficiency for SBLOCA's. Included in these models is a stable, implicit solution method which has provided a significant reduction in CPU requirements while maintaining satisfactory analyses results. The new models for RETRAN-03 and results of small break analyses using RETRAN-03 are briefly discussed in this paper.     

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.     

IIST small break LOCA experiments with passive core cooling injection

The purpose of this study is to evaluate the performance of a passive core cooling system (PCCS) with passive injection during the cold-leg small break loss-of-coolant accidents (SBLOCAs) experiments conducted at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. Four tests were performed simulating break sizes of 0.2–2% (approximately corresponding to 1.25–4″ breaks for a referenced nuclear power plant) at cold-leg for assessing the PCCS capability in accident management. The key thermal–hydraulic phenomena to core heat removal for PCCS are observed and discussed. The experimental results show that the PCCS has successfully provided a continuous removal of core heat and a long term core cooling can be reached for all cases of SBLOCA.     

A 10% cold leg break test at ROSA-IV large scale test facility

The Large Scale Test Facility (LSTF) in the ROSA-IV Program is an integral test facility for investigation of pressurized water reactor (PWR) thermal-hydraulic behavior during small break loss-of-coolant accidents (SBLOCAs) and operational transients.A 10% cold leg break test was conducted in the facility shakedown phase to assess and confirm the facility capability and to collect code assessment data. The test conditions, test procedures and test results are described. The test results are compared with a pretest analysis obtained using RELAP5/MOD1 Cy18.     

Analysis of an RPV upper head SBLOCA at the ROSA facility using TRACE

Inspections of existing nuclear power plants have pointed out the possibility that small break loss-of-coolant accidents (SBLOCAs) could be initiated by a small break located in the upper head (UH) of the reactor pressure vessel (RPV). Such type of breaks has been the subject of investigation in some of the tests carried out in the framework of the OECD/NEA ROSA test program for safety research and safety assessment of light water reactors. The ROSA/LSTF test facility simulates a Westinghouse design PWR with a four-loop configuration and 3423 MWth. Areas, volumes and power are scaled down by a factor of 1:48 while the elevations are kept at full height. Only two loops, sized to conserve the volume scaling (2:48), are simulated. The present paper is focused on test 6-1 that simulated a RPV upper head SBLOCA with a break size equivalent to 1.9% cold leg break. The experiment assumes a total failure of the high pressure injection system (HPIS) and a loss of off-site power concurrent with the scram. The main purpose of the present study is the assessment of the capabilities of the best estimate system code, TRACE, to reproduce and understand the physical phenomena involved in this type of SBLOCA scenarios. Special attention was dedicated to the modelling of the leakage flows, necessary to correctly simulate the distribution of the water inventory in the primary side. In addition, the particular location of the break in test 6-1 allows the verification of the chocked flow model in the same way as for a separate-effect test.     

高通量工程试验堆出口管道焊缝破口频率计算

近年来,国内外进行多项研究堆概率安全分析,其中管道破口导致的失水事故是堆芯损坏的重要风险来源。本文参考管道破口计算程序PRAISE(Piping Reliability Analysis Including Seismic Events)方法,选取压力壳型研究堆——高通量工程试验堆(High Flux Engineering Test Reactor,HFETR)的运行工况,对其反应堆冷却剂出口管道的焊缝进行分析,得到运行中该处焊缝发生各类破口的频率。     

PACTEL integral test facility – Description of versatile applications

Regardless of the large number of thermal hydraulic experiments conducted with many different facilities, the need for good quality data from integral test facilities has not yet reached saturation. The parallel channel test loop (PACTEL) facility is one of the largest facilities of its kind. It was originally designed to model the thermal-hydraulic behavior of VVER-440 type pressurized water reactors (PWRs) currently used in Finland. Nevertheless, the PACTEL facility has served also in many other purposes than for VVERs only. The facility has been modified on a case-by-case basis according to the needs in configuration and positioning of auxiliary equipment. The newest plan is to modify PACTEL to be the PWR PACTEL, a facility with vertical steam generators for EPR applications topical in Finland. This paper describes the versatile use of the PACTEL facility for a large spectrum of thermal hydraulic research. The PACTEL facility is ideal for investigating planned recovery procedures during accidents and operational transients. For this purpose experimental series among others on small break loss-of-coolant accidents (SBLOCA), primary-to-secondary leakages (PRISE), and on anticipated transients without scram (ATWS) have been carried out. The PACTEL natural circulation experiment with stepwise coolant inventory reduction formed the basis for the OECD International Standard Problem (ISP-33). In addition, many other one-phase and two-phase natural circulation tests have been executed.     

Analysis of loss-of-coolant and loss-of-flow accidents in the first wall cooling system of NET/ITER

This paper presents the thermal-hydraulic analysis of potential accidents in the first wall cooling system of the Next European Torus or the International Thermonuclear Experimental Reactor. Three ex-vessel loss-of-coolant accidents, two in-vessel loss-of-coolant accidents, and three loss-of-flow accidents have been analyzed using the thermal-hydraulic system analysis code RELAP5/MOD3. The analyses deal with the transient thermal-hydraulic behavior inside the cooling systems and the temperature development inside the nuclear components during these accidents. The analysis of the different accident scenarios has been performed without operation of emergency cooling systems. The results of the analyses indicate that a loss of forced coolant flow through the first wall rapidly causes dryout in the first wall cooling pipes. Following dryout, melting in the first wall starts within about 130 s in case of ongoing plasma burning. In case of large break LOCAs and ongoing plasma burning, melting in the first wall starts about 90 s after accident initiation.     

Experimental Study on Heat Removal during Cold Leg Small Break LOCAs in PWRs

Experiments which simulated small break loss-of-coolant accidents (SBLOCAs) resulting from 2.1–0.13% break in the cold leg of a PWR were conducted with an apparatus of 1/270 scale in volume. In the large break size case, the decay heat was mainly removed by the break flow and in the case of a small break, the steam generator played an important role. In this case, thermal hydraulic behaviors such as natural circulation and reflux condensation cooling were important during the transient. Depressurization in the secondary system due to bleeding steam from the steam generator by an operator action was so effective to make the accident to come to an end. The operation to depressurize the secondary system was also efficient to rewet the core which had been uncovered due to a loop seal formation in a cross-over leg.

No effects of initial 200 ppm dissolved gas in the coolant were observed on the cooling performance of the steam generator. It was considered that it was because the gas which came from the coolant into the steam during the depressurization transient did not remain in the tubes of the steam generator.     

FRAP-T6: A computer code for the transient analysis of oxide fuel rods

The FRAP-T6 computer code was developed to model the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to large break loss-of-coolant accidents. The code models all of the thermal, structural, and chemical phenomena needed for the complete evaluation of light water reactor fuel rod performance. The code was developed using rigorous quality assurance procedures and a large assessment data base. The results of assessment show that the code accurately models the response of light water reactor fuel rods.     

Power-operated relief valve stuck-open accident and recovery scenarios in the Institute of Nuclear Energy Research integral system test facility

Four scaled small break loss-of-coolant accident (LOCA) tests simulating the pressurizer power-operated relief valves (PORVs) stuck-open accidents and the recovery actions in a pressurized water reactor (PWR) were performed at the Institute of Nuclear Energy Research (INER) integral system test (IIST) facility. The objectives of this study are to verify the effectiveness of emergency operating procedure (EOP) and emergency core cooling system (ECCS) on reactor safety. The break sizes were volumetrically scaled down based on one and all three fully-opened PORVs which is equivalent to 0.23% and 0.69% hot leg flow area of the reference plant. The experimental results indicate that in case of high pressure injection (HPI) system failure, the rapid depressurization of the steam generators is proved to be an effective way in the depressurization of the reactor coolant system and the core cooling. In contrast, if only one HPI charging pump operates normally, which injected half (or minimum) flow rate of normal cooling water, the core cooling can be adequately provided without operating the secondary bleeding during PORV stuck-open transient. This paper also presents the scaling methods for the reduced-height, reduced-pressure (RHRP) IIST facility and the test conditions. The validity of the present scaling methodology is confirmed by the results from previous IIST counterpart tests and comparison of the present results with those of the tests performed at the full-height, full-pressure(FHFP) stuck-open tests.     

Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TM1–2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating.     

Analysis of severe core damage accident progression for the heavy water reactor

In this study,the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code.The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems.The progressions of severe accident included a set of failed safety systems normally operated at full power,and initiative events led to primary heat transport system inventory blow-down or boil off.The core heat-up and melting,steam generator response,fuel channel and calandria vessel failure were analyzed.The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault.     

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.     

Verification of the two-phase stratified-flow model in by separate effect tests

The code which is being developed by the Gesellschaft für Anlagen- und Rcaktorsicherheit (GRS) mbH is intended to cover, by means of a single code, the entire spectrum of loss-of-coolant and transient accidents in pressurized and boiling water reactors. The actual version Mod 1.1-Cycle A has a five-equation two-phase model based on the conservation laws for liquid mass, liquid energy, vapor energy and overall momentum. The relative velocity between liquid and vapor is determined by a full-range drift-flux model for two-phase flow in horizontal and vertical pipes. The verification of this drift-flux model is carried out by both large-scale experiments and single-effect tests. The single-effect test ECTHOR investigates stratified flow during the clearance of a water-filled loop seal by a forced air flow through the loop. ECTHOR is a French test for the consideration of two-phase flow regimes in pipes for the development of the codes. The experiments are dedicated to investigating typical two-phase flow during small break loss of coolant accidents (LOCA) in pressurized water reactors (PWR).As a measure, the remaining water level in the loop is determined as a function of the air flow rate. For the verification, a comparison between and computations, on the one hand, and experiments on the other hand is carried out. The results compare very well to each other. Test runs on different numerical grids show convergence to an asymptotic limit with increasing grid refinement.     

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.     

Reliability of Core Exit Thermocouple for Accident Management Action during SBLOCA and Abnormal Transient Tests at ROSA/LSTF

Presented are experimental results on the general performances of core exit thermocouples (CETs) to detect core overheat for accident management (AM) action by using the Large-Scale Test Facility (LSTF) of the ROSA Program of the Japan Atomic Energy Agency. The LSTF is a full-height, full-pressure, and 1/48-volumetric-scaled model of a 4-loop pressurized water reactor (PWR). This study was motivated by a significant delay in the time and temperature rise of the CETs from core heat-up during a vessel top head small break loss-of-coolant accident (SBLOCA) test. A certain delay in time and temperature rise of the CETs was also observed in various SBLOCA and abnormal transient tests. Such CET performances are derived from thirteen LSTF tests as follows: (1) general CET performances are obtained in the form of equations including cases under limited influences of water fall-back from hot legs, (2) the major reason for the delay is the interaction of three-dimensional steam flows with low-temperature structures in and around the core and core exit, (3) break location was insignificant except for the PV top and bottom break cases, and (4) CET superheat is suitable for AM instead of the temperature value for significantly high or low pressure transients. The applicability of the results to PWR is discussed further.     

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.     

Numerical simulation of natural circulation experiments conducted at the IIST facility

Natural circulation plays an important role in long-term cooling of pressurized water reactors (PWRs) under small break loss-of-coolant accidents. Recently, natural circulation experiments have been conducted at the Institute of Nuclear Energy Research integral system test (IIST) facility, which is used to simulate the Westinghouse three-loop Maanshan PWR. A numerical simulation is presented to investigate the natural circulation phenomena of the IIST facility with the RELAP5/MOD3 code. The calculated results are in good agreement with the experimental data of the single-phase natural circulation both quantitatively and qualitatively. The influences of power level and system pressure on natural circulation can also be predicted by the current model. Based on the two-phase natural circulation data, the calculated flow rate history is similar to that obtained from the experiment.     

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.