Analysis of an ADS spurious opening event at a BWR/6 by means of the TRACE code

The paper presents the results of a post-event analysis of a spurious opening of 8 relief valves of the automatic depressurization system (ADS) occurred in a BWR/6. The opening of the relief valves results in a fast depressurization (pressure blow down) of the primary system which might lead to significant dynamic loads on the RPV and associated internals. In addition, the RPV level swelling caused by the fast depressurization might lead to undesired water carry-over into the steam line and through the safety relief valves (SRVs). Therefore, the transient needs to be characterized in terms of evolution of pressure, temperature and fluid distribution in the system. This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the plant data.     

Soil-structure interaction for transient loads due to safety relief valve discharges

Dynamic responses of BWR Mark II containment structures subjected to axisymmetric transient pressure loadings due to simultaneous safety relief valve discharges were investigated using finite element analysis, including the soil-structure interaction effect. To properly consider the soil-structure interaction effect, a simplified lumped parameter foundation model and an axisymmetric finite element foundation model with viscous boundary impedance are used. Analytical results are presented to demonstrate the effectiveness of the simplified foundation model and to exhibit the dynamic response behavior of the structure as the transient loading frequency and the foundation rigidity vary. The impact of the dynamic structural response due to this type of loading on the equipment design is also discussed.     

Boiling water reactor stability revisited: The effects of flashing

There are numerous investigations of two-phase flow stability with particular emphasis to BWR stability; these have become increasingly sophisticated and complete over the years. The basic features of a new development and frequency-domain code capable of considering all the channels (bundles) in a BWR, flashing of the coolant at low pressure, full coupling with 3D, two-group neutronics, etc. are described. The basic thermal-hydraulic model is used to study the effects of flashing on stability in a BWR-like channel. The behavior of the channel is highly dynamic. Contrary to what could have been intuitively guessed, the effect of flashing is stabilizing; the reasons and mechanisms leading to this are discussed.     

Simplified distributed parameters BWR dynamic model for transient and stability analysis

This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR.     

The steamhammer problem: dynamic shock loading of critical reactor piping systems

In the course of both pre-operational testing and power operation of commercial nuclear power plants, relatively large amplitude transient vibrations of steam piping systems have been experienced with damage to the piping supports in at least one recent case. These transient vibrations result from ‘steamhammer’ or dynamic shock loading induced by pressure and momentum transient conditions generated in the piping by sudden changes to the flow conditions, such as are produced by sudden valve opening or closure. In particular, vibrations have been experienced in by-pass and discharge lines as a result of relief valve discharge, and in main steam lines as a result of sudden main stop valve closure. Piping in both BWR and PWR reactor systems has been found to be susceptible to these conditions.This paper is concerned with the evaluation of the pressure and momentum transients resulting from sudden valve operation, and the determination of the dynamic response of the piping to the induced transient loading. The characteristics of the transient conditions existing immediately following both sudden valve opening and closure as encountered in BWR and PWR plants are discussed. The procedures used to calculate the transient time history functions are outlined. The derivation of the loading induced in the piping by the pressure and momentum transients is discussed and the determination of the dynamic response of the piping is presented. The procedures described in the paper are illustrated by actual examples from BWR and PWR plants, and the significance of steamhammer effects relative to other loading conditions is discussed.     

Study on the minimum number of SRVs required for emergency depressurization in EOP of BWR power plants

The minimum number of safety/relief valves (SRVs) required for emergency depressurization (MNSRED) is an important reactor pressure vessel (RPV) parameter described in the appendix C of BWR Owners’ Group, Emergency Procedure Guidelines and Severe Accident Guidelines (BWROG, EPG/SAGs). It is directly related to the emergency depressurization if there are less SRVs available. MNSRED is always coupled with decay heat removal pressure (DHRP) in emergency depressurization. DHRP is used as the depressurization low-pressure end to ensure the decay heat can be removed.The purpose of this paper is to discuss their mathematical models in BWROG EPG/SAGs. After study, the vague concepts existed in these parameters are pointed out and clarified. The variables inappropriately used through their evaluations are revised. With improved modifications, the MNSRED turns out smaller than its original value and DHRP is larger than its previous one. These results affect the heat removal capability and the operation margin regarding emergency depressurization. At the end, the revised DHRP value is verified by MAAP5 code with the same plant status for the Kuosheng power station of Taiwan Power Company.     

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.     

Transient Behaviors of Gas in Thermal Insulation Media at Rapid Depressurization

Thermal insulation systems are often used to high-temperature gas cooled reactors.

In this paper, transient behavior of the gas flow in thermal insulation media was systematically analyzed in the case of sudden decrease in a coolant pressure. The structure of these systems is capable of substituting for the simplified model that the gas in a containment filled with porous insulation media exhausts from a balancing hole to main flow duct. In the case of rapid depressurization, these behaviors were calculated by using equations of continuity and motion based on the Darcy flow model with area ratio of a balancing hole and permeability of porous insulation media as parameters. The transient pressure difference of a hot liner wall between the main flow helium and the insulation layer induced at depressurization was numerically analyzed. It was revealed from analytical results that transient behaviors were strongly influenced by prescribed parameters, and experimental results can be explained by using the analytical model.     

超临界流体的泄压过程研究

超临界水堆(SCWR)的LOCA研究是安全分析的重点和难点,其中压力容器的喷放泄压过程的研究至关重要。本文通过对反应堆压力容器进行简化,建立了简单容器喷放的数学物理模型,开发了超临界流体的喷放瞬态计算程序。将该程序的计算结果与超临界二氧化碳的泄压喷放过程的实验数据进行了比较,计算值与实验结果吻合良好,验证了模型的正确性。运用该验证后的程序对超临界水的容器喷放过程进行了深入研究和分析,分析了不同初始条件、破口面积及加热功率等对泄压过程瞬态特性的影响。结果表明,本文建立的简单容器模型能模拟从超临界到亚临界压力的喷放泄压过程。计算结果可为超临界水堆的LOCA分析提供理论基础。     

Effect of depressurization on reactor vessel inventory in the absence of ECCS injection

A series of tests were performed to evaluate inventory depletion as a reactor vessel undergoes depressurization in the absence of any emergency core coolant system injection (ECCS). These tests were carried out in a scaled representation of a reactor vessel which was initially filled with saturated water up to the elevation of the hot legs. Depressurization valves installed on take-off lines from the hot legs were opened and level swell ensued in the reactor vessel initiating a two-phase blowdown. This was followed by subsequent single-phase discharge transient which in some cases led to core uncovery. A combined model encompassing the two-phase and single-phase discharge portions of the transient is proposed. The inventory-versus-pressure traces obtained from the model compare well with the experimental results. These traces are discussed as bounding trajectories for a large class of small break loss of coolant accident (LOCA) transients which otherwise must be considered individually.     

稳压器卸压阀卸压效果研究

基于国际上模拟严重事故瞬态过程最详细的机理性程序SCDAP/RELAP5/MOD3.1,主要分析研究了核电站未紧急停堆的预期瞬变(ATWS)初因(失去主给水、失去厂外电和控制棒失控提升)叠加辅助给水失效导致的堆芯熔化严重事故进程,并验证阻止ATWS导致堆芯熔化进程的一次侧卸压缓解措施的充分性和有效性.计算分析结果显示,一列稳压器卸压阀不足以充分降低一回路压力,压力仍然停留在10MPa以上,存在很大高压熔堆的风险.增加一列卸压阀可把一回路压力降低到3MPa左右,安注系统得以投入,及时有效地阻止堆芯熔化进程,降低了高压熔堆风险.分析结果还显示高压安注系统的投入对一回路卸压具有重要影响.     

PWR Cold-Leg Small-Break LOCA with Total HPI Failure

Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5–10% of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insufficient for intermediate break areas to maintain adequate core cooling. It is also shown that there might be possibility of core dryout after ACC injection and before LPI injection for break areas less than approximately 2.5%.     

Evaluation of Delayed Neutron Fraction β for Thermal Fission of U-235 Based on Integral Experiments Using SHE

Countercurrent gas-liquid flow is theoretically and experimentally evaluated for a boiling system simulating a BWR core. In a single channel, flow patterns are determined from the mass balance equations and pressure drop under steady state conditions is calculated for each flow pattern using a drift flux model, where the distribution parameter and drift velocity are correlated as functions of void fraction and hydraulic diameter from void fraction data. The calculated pressure drop shows a similar trend to that of the data for the effects of bypass leak flow rate and heater power. Countercurrent behavior in three boiling channels under slow transient conditions is also predicted from the single channel characteristics and close agreement is obtained between the predicted and experimental results. The results show that steam up-flow or cocurrent up-flow easily occurs in a channel with low pressure drop, namely, with a large entry orifice, high power or low bypass leak flow rate.     

Numerical Study of Depressurization Rate during Blowdown Based on Lumped Model Analysis

To study the thermodynamic aspects of blowdown, the depressurization rate equation has been numerically solved. The equation, derived from macroscopic mass and energy balances in the pressure vessel, consisted of the energy and volumetric discharges terms multiplied by the decrease rate of residual coolant. By applying a dimensional analysis, dimensionless equations were obtained together with dimensionless parameters of blowdown. Blowdown calculations starting at typical BWR operating conditions indicated that the decrease rate of coolant increased for the liquid and two-phase mixture, and decreased for the vapor discharge. Further, the energy discharge term made a larger contribution to the depressurization rate in the case of vapor escape, while the volumetric discharge term did so in the case of liquid and two-phase mixture escape blowdowns. In the lumped model analyses, the averaged specific enthalpy and entropy of the residual coolant increased for the liquid discharge, remained almost constant for the two-phase mixture discharge, and decreased for the vapor discharge blowdown.     

The prestressed cast iron pressure vessel (PCIV): Its applicability for gas- and water-cooled nuclear power reactors and for burst protections

The field of applicability of prestressed cast iron pressure vessels in the nuclear power industry is described. Principles of design and construction are discussed. Theoretical analyses are corroborated by results obtained from model testing. Pod boiler versions of the PCIV are introduced for two HTRs. The design of a PCIV for a BWR is proposed. Two concepts of a prestressed cast iron burst protection for steel pressure vessels of a BWR are presented. The advantages of prestressed cast iron structures are listed.     

超临界水实验回路热工水力分析

作为超临界水堆失水事故分析的关键现象,跨临界过程(即超临界水堆的压力从超临界状态降到次临界状态22.1 MPa以下)受到国内外的关注。上海交通大学的超临界流体多功能实验回路(SWAMUP)计划对这一泄压过程进行实验研究。为确保该实验装置在实验过程中的安全性能,采用系统程序ATHLET-SC对该实验回路进行预计算分析,主要针对该系统在泄压跨临界过程中的热工水力参数,包括系统压力、冷却剂流量、加热棒壁面温度等展开计算,并讨论一些重要参数如泄压速度、加热棒加热功率等对计算结果的影响。计算结果表明,修改后的ATHLET-SC程序可模拟跨临界瞬态过程,在实验过程中,加热棒壁面温度不会超过设计上限温度,然而,回路中换热器的内外最高压差将会达6 MPa,这一点需在实验中特别考虑。     

Analyses of pressure perturbation events in boiling water reactor

In order to design more stable and safer core configurations, experimental and theoretical studies about BWR (Boiling Water Reactor) instability have been performed to characterize the phenomenon and to predict the conditions for its occurrence. The instabilities can be caused by interdependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In this work, the RELAP5-MOD3.3 thermal-hydraulic system code and the PARCS-2.4 3D neutron kinetic code were coupled to simulate BWR transients. The pressure perturbation is considered in order to study in detail this type of transient. Two different algorithms developed at the University of Pisa were used to calculate the Decay Ratio (DR) and the natural frequency (NF) from the power oscillation signals obtained from the transient calculations. The validation of a code model set up for the Peach Bottom-2 BWR plant is performed against Low-Flow Stability Tests (LFST). The four series of Stability Tests were performed at Peach Bottom Unit 2 in 1977 at the end of cycle 2 in order to measure the reactor core stability margins at the limiting conditions used in design and safety analysis.     

A blind simulation of RD-14M small-break LOCA experiments using CATHENA code

The results of two Small Break Loss of Coolant Accident (SBLOCA) experiments in RD-14M test facility and their predictions by CATHENA code, which is used to analyze postulated events in CANDU reactors, are compared in this paper. Two specific SBLOCA experiments selected for the CATHENA code predictions are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow-down.These blind simulations demonstrate that CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. The CATHENA predictions of fuel sheath temperatures for test B9006 are in much better agreement with the test measurements than those for test B9802, because CATHENA code could not capture the oscillatory behavior of channel flow and consequently sheath temperature when local fuel sheath surface was being intermittently dried and rewet under near stagnant flow and full power conditions of test B9802. Therefore, it is concluded that further works are required to appropriately predict the sheath temperature spike and its fluctuation during the transient of test B9802 where the critical heat flux and post-dryout heat transfer are important governing phenomena.     

Detailed evaluation of coolant injection into the reactor vessel with RCS depressurization for high pressure sequences

A coolant injection into the reactor vessel with depressurization of the reactor coolant system (RCS) has been evaluated as part of the evaluation for a strategy of the severe accident management guidance (SAMG). Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feedwater (LOFW) accident in Optimized Power Reactor (OPR)1000 have been analyzed by using the SCDAP/RELAP5 computer code. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 s with indirect RCS depressurization by using one condenser dump valve (CDV) at 6  min after implementation of the SAMG prevents reactor vessel failure for the small break LOCA without SI. In this case, only one train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent reactor vessel failure. Only one train operation of the HPSI at 20,208 s with direct RCS depressurization by using two SDS valves at 40 min after an initial opening of the safety relief valve (SRV) prevents reactor vessel failure for the total LOFW.     

A simplified model for the dynamics of a BWR

A model has been developed to derive the dynamic characteristics of a BWR with natural circulation. The model is based on the basic physical processes that govern reactor dynamics. The actual values for the model parameters are estimated from experimental and theoretical data. The model enables the computation of transfer functions of reactivity and steam flow to power and pressure. The sensitivity of these transfer functions to changes in model parameters is discussed.