Critical crack depth diagram of reactor vessel for pressurized thermal shock

During the operation of a pressurized water reactor, a certain type of transients could induce rapid cooldown of the reactor pressure vessel (RPV) with relatively high or increasing system pressure. This induces a high tensile stress at the inner surface of the RPV, which is called the pressurized thermal shock (PTS). The structural integrity of the RPV during PTS should be evaluated assuming the existence of a flaw at the vessel. For the quantitative evaluation of the vessel failure risk associated with PTS, the probabilistic fracture mechanics (PFM) analysis technique has been widely used. But along with PFM analysis, deterministic analysis is also required to determine the critical time interval in the transient during which mitigating action can be effective. In this study, therefore, the procedure for the deterministic fracture mechanics analysis of RPV during PTS is investigated using the critical crack depth diagram and the computer program to generate it is developed. Four transients of typical PTS, steam generator tube rupture, small break loss of coolant accident and steam line break are analyzed, and their response characteristics such as critical crack depth and critical time interval from the initiation of the transient are investigated.     

承压热冲击下压力容器断裂力学分析

依据美国核管会（NRC）最新法规要求和研究进展,阐述了压水堆核电厂反应堆压力容器（RPV）承压热冲击（PTS）最新评估方法。基于热工水力系统程序RELAP5和有限元分析软件ANSYS,针对某传统二代压水堆核电厂模拟在PTS典型瞬态过程下热工响应行为及压力容器模型断裂力学分析,并评估不同瞬态的危险性及其随压力容器材料脆性的变化。分析表明：表面裂纹和靠近内壁面的埋藏裂纹比深埋裂纹更易发生开裂;同等条件下轴向裂纹较环向裂纹更易开裂,且大中破口事故下轴向裂纹远较环向裂纹更易贯穿壁厚。     

Analysis of the Leibstadt power plant condensate and feedwater systems during selected operational transients

This paper presents a selection of plant analyses that were carried out by PSI in support of the Leibstadt Nuclear Power Plant (Swiss boiling water reactor). The analyses were performed as part of a collaboration between Leibstadt and PSI, to help resolve some operational problems that were experienced during the power uprate beginning in 1998. The issues under investigation were related to the behavior of the condensate and feedwater systems during transients initiated by a turbine trip, load rejection and a single feedwater pump trip, all of which increased the risk of an inadvertent reactor shutdown by reaching reactor pressure vessel water level limits. The possibility of a reactor shutdown was related to perturbations in the feedwater flow caused by transitory pump cavitation of the feedwater pumps, due to a rapid depressurization in the feedwater tank. In addition to a direct analysis of plant measurement provided by Leibstadt, steady-state and transient simulations of the events were performed at PSI using the system codes TRAC-BF1 and TRACE. Through a combination of the analysis of the plant measurements, the code simulations and an analysis of the whole plant behavior using the Leibstadt plant-simulator appropriate modifications of the plant hardware, control system and operational set points were proposed. The implementation and success of these changes were verified by a number of plant tests. Finally, the original designed plant capability not to shutdown during the aforementioned transients was demonstrated.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

承压热冲击下材料特性对反应堆压力容器结构完整性的影响

介绍了承压热冲击(PTS)分析的背景和研究现状,阐述了基于确定性断裂力学的反应堆压力容器(RPV)结构完整性分析方法.分析了材料性能模式(线弹性和弹塑性)和辐照效应对PTS下RPV结构完整性的影响.     

Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

Vessel behavior following a through-wall crack

A fracture mechanics model has been developed to predict the behavior of a reactor pressure vessel following the occurrence of a through-wall crack during a pressurized thermal shock event. This study has been coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory. The fracture mechanics model uses as inputs the critical transients and probabilities of through-wall cracks from the IPTS Program. The model has been applied to predict the modes of failure for plant specific vessel characteristics. A Monte Carlo type of computer code has been written to predict the probabilities of alternate failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. The computer code also calculates the crack driving force as a function of the crack length and the internal pressure for critical times during the transient. The fracture mechanics model has been applied in calculations that simulate the Oconee-1 reactor pressure vessel. The model predicted that about 50% of the through-wall axial cracks will turn and follow a circumferential weld giving a potential for missiles. Missile arrest calculations predict that vertical as well as all potential horizontal missiles will be arrested and will be confined to the vessel enclosure cavity. In future work, plant specific analyses will be continued with calculations that simulate Calvert Cliff-1 and H.B. Robinson-2 reactor vessels.     

反应堆压力容器承压热冲击中的PSA方法研究

在瞬态过程中,当处于承压状态下的反应堆压力容器(RPV)的内表面被快速冷却时,即为承压热冲击(PTS)。由此,反应堆压力容器可能出现贯穿裂纹而失效。为分析PTS事件导致RPV出现裂纹的频率,需要进行概率安全评价(PSA)。通过PSA模型确定可能引起PTS的事件序列,并结合这些序列的热工水力分析结果,为PTS概率断裂力学分析提供支持。     

压力温度限值曲线计算与比较分析

核电厂在启停堆过程中必须将压力和温度控制在限值范围内,即压力温度限值曲线(P-T曲线)所允许的范围内,以防止反应堆压力容器发生脆性开裂。以法国的RCCM规范、美国的ASME规范和我国的核行业标准EJ/T918为对象,对比分析P-T曲线各自的计算方法,讨论了所采用的保守假设对计算结果的影响。研究表明,选取材料静态断裂韧性KIC计算P-T曲线将会增加核电厂启停堆过程中的操控空间;与最新版国外规范比较,我国行业标准EJ/T918的计算结果显得过于保守。     

The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using Lungmen startup test data.     

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.     

反应堆压力容器承压热冲击分析研究

依据RCC-M规范和美国NRC 10CFR50.61,对存在假想裂纹的反应堆压力容器堆芯带区进行承压热冲击分析研究.计算核电厂寿期末的基准温度,并采用承压热冲击筛选准则进行评定;计算了承压热冲击瞬态作用下裂纹尖端的应力强度因子,并按RCC-M规范进行评定.     

1991 National Symposium on Atomic Energy,Tokyo, Japan

The probabilistic fracture mechanics analysis code, PROFMAC-11, was developed and benchmarked for postulated pressurized thermal shock (PTS) events of pressurized water reactor pressure vessel integrity issues by participating the PTS benchmarking activities organized and managed jointly by U.S. Nuclear Regulatory Commission (USNRC) and Electric Power Research Institute (EPRI). Three different sets of benchmarking problems were assigned to participants for the studies in sequence. The initial and second sets were preparatory to the third set. Results from the sequential benchmarking activities were that for axial flaw analyses probabilities of crack initiation were identical to those for vessel failure and the scatter of probabilities among four participating codes was small, while for circumferential flaw analyses probabilities varied widely. It was eventually verified that the same probabilities were achievable by unifying procedures for simulating material properties. Performance of the PROFMAC-II code was excellent in terms of consistency and adequacy among the four codes and the eight organizations which took part in the studies. Several issues relevant to analysis techniques and input variables were pointed out for further improvement of the codes.     

The HTTR reactor pressure vessel and its integrity against a PTS event

Before manufacturing the real steel to be used in the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) the vessel manufacturer and materials supplier made a sample steel by the same procedure as for the real steel (2.25Cr-1 Mo) and conducted many tests to obtain material strength data for its base and weld metals. The test results showed that the sample steel satisfied the HTTR design requirements. Vessel cooling panels are set on the inner surface of the biological shielding concrete around the RPV, and are circulated with cooling water at 0.5 MPa and 40°C to cool the shielding concrete during normal operation of the reactor. By supposing that the cooling panel breakes and the water discharges to the RPV outer surface heated at 400°C, the stress distribution generated in the vessel wall by a pressurized thermal shock (PTS) event can be calculated using a finite element method code. This paper describes some of the results obtained from the material testing of the sample steel and the estimated result using the scheme developed for a light water reactor pressure vessel, to clarify the integrity of the HTTR-RPV under a PTS event.     

Three-dimensional simulation of the coupled convective, conductive, and radiative heat transfer during decay heat removal in an HTR

A three-dimensional model has been constructed to simulate the passive heat removal in a modular prismatic-block high temperature reactor during a loss of active cooling accident. This model, developed using the STAR-CD general computational fluid dynamics code, solves the combined conductive, convective, and radiative heat transfer within a 30° section of the core and reactor vessel. To accommodate the different spatial scales, it uses homogeneous equivalent media to represent the coolant flow and the prismatic fuel blocks. A customized procedure that manages solving alternatively the dynamic and thermal fields permits the computation of very long transients, which typically are performed for 100 or more hours of simulated time.The global methodology and specific modeling procedures are explained, and key points of the CFD analysis are highlighted. Next, the results of several calculations are presented, and the physical phenomena represented are described. Two commonly investigated loss of active cooling scenarios are considered: depressurized conduction cooldown and pressurized conduction cooldown. The results for these two scenarios are compared to assess the effect of heat transfer via internal natural convection – which is negligible during the depressurized event – on the thermal behavior of the system. In addition, the evolution of the natural convection flow through the core and in the annular spaces is examined and discussed.     

Results of probabilistic fracture mechanics analyses for pressurized thermal shock transients in the reactor pressure vessel of BIBLIS-B

Probabilistic fracture mechanics investigations of the contribution of pressurized thermal shock transients to reactor pressure vessel failure probability of the reference plant for the German reactor safety study phase B, BIBLIS-B, are presented. The applied methods and the calculation model are discussed. The most important result of parametric analyses is that the postulated flaw distribution in the vessel has a dominant influence on the calculated conditional failure probability. With regard to the transient behavior the results show, that the temperature drop induced by the thermal shock has great influence on the conditional failure probability, whereas the decay rate of the temperature change has minor influence.     

Probabilistic fracture mechanics analysis of nuclear structural components: a review of recent Japanese activities

This paper describes a review of recent Japanese activities on probabilistic fracture mechanics (PFM) analyses. Japan Atomic Energy Research Institute (JAERI) has sponsored research committees on PFM organized by Japan Society of Mechanical Engineers (JSME) and Japan Welding Engineering Society (JWES) for more than 10 years. The purpose of the continuous activity is to establish standard procedures for evaluating failure probabilities of Japanese nuclear structural components such as PV&P and steam generator tube, combining the state-of-the-art knowledge on structural integrity of nuclear structural components and modern computer technology such as parallel processing. This paper shows two topics of the newest results of JWES committee, PFM analysis of aged reactor pressure vessel considering embedded cracks and PFM analysis of piping considering seismic loading, and one topic by JAERI itself, development of PTS analysis code for transient loading (PASCAL).