Estimates of brittle fracture probability using reactor surveillance capsule data

This work is part of an expanding effort to construct sound methodology for calculating estimates of brittle fracture of pressure vessels. These estimates are generated from material data on the class of steels being used and from calculated operating characteristics, and involve the structuring of K_IC and ΔRT_NDT as random variables. Three sources of variability are inherent in material data of the type involved here: (a) measurement variability, (b) variability arising from local inhomogeneities in a plate, and (c) variability from gross differences among plates. Generic data reflect all three, while vessel-specific data limit the third to those plates that are actually used in fabricating a given vessel. We have devised a procedure that combines the data sets with weights that reflect the variance composition and would agree with the limiting cases above should one of them be true.Vessel specific data come from two sources:

1. (1) preoperational tests — these are Charpy V-notch tests from the exact steel plates which are used in fabricating the pressure vessel. Before a plate is used in vessel fabrication, specimens are machined from it and Charpy tests run. Only those plates with acceptable Charpy values are used.

2. (2) Surveillance capsules — in order periodically to monitor the characteristics of an operating pressure vessel, surveillance capsules are placed in the reactor prior to start-up. Each capsule contains specimens machined from the steel plates used in fabricating the vessel. These capsules are removed periodically during scheduled plant shutdowns and tensile, Charpy and fracture toughness tests are run. These are very valuable data since they have been irradiated in the specific reactor of interest.

In this paper we present a method for estimating the brittle fracture probability using the HSST and vessel-specific data. The method is general enough so that data from all surveillance capsules removed from a vessel to date can be included for the purpose of updating estimates of brittle rupture probability at the instances of scheduled shutdowns.     

A simplified refueling scheme for nuclear heating reactor

The nuclear heating reactor is a clean energy resource with high reliability and high economic benefits. Its basic design characteristics are different from those of big reactors. Some features, such as in-pressure vessel spent fuel storage, long refueling period, etc., provide the possibility to simplify the refueling system for economic purposes. After being stored in a pressure vessel for about 15 years, the spent fuel assemblies, with very low radioactivity and decay heat capacity, may be removed from the reactor pressure vessel to a storage pool by a simplified system including a shielded flask, reactor building crane, and some auxiliary tools. It is demonstrated that this ‘dry-method’ refueling scheme is safe and reliable.     

WWER-1000 base metal reference steel and its characterisation

The surveillance programmes of western power reactors include, in many cases, standard reference materials in addition to actual pressure vessel steels. These are specimens cut from standard steel plates (Heavy Section Steel Technology, JRQ, etc.) that are similar in composition and heat treatment to the base material in the respective reactor pressure vessels, and are supposed to serve as a monitor by comparing the radiation embrittlement of the plant-specific material to the reference material, and to detect anomalies in the radiation environment of the surveillance capsules.A correlation monitor material for the eastern WWER-1000 (similar as the JRQ for western reactors) is needed in order to determine the reliability of accelerated data for the validation of reactor pressure vessel surveillance data. Reference materials should be well characterised in terms of irradiation behaviour (transition temperature shift, non-destructive signal, etc.). The magnitude of the sensitivity to irradiation for this material should be measurable for the selected exposures. In this subject the IAEA is launching a new co-ordinated research programme. Material is already manufactured, and the JRC-IE has become its custodian. A detailed plan for characterisation of the reference steel is set up, including irradiation conditions, post-irradiation testing techniques and implementation plan. It is expected the participation of several research institutes worldwide in a round robin, which will allow a better comprehension of WWER-1000 steel's behaviour and will be considered as a benchmarking between different laboratories.The JRC-Institute for Energy in collaboration with the Russian Research Centre – Kurchatov Institute is performing the “as received” material characterisation by both destructive methods and non-destructive techniques.The non-destructive techniques used at the JRC-IE premises are novel methods specially developed for non-destructive assessment of the embrittlement state of materials, as the STEAM method and the measurement of magnetic properties. The STEAM technique (Seebeck and Thomson effects on aged material), is based on the measurement of the Seebeck coefficient. The magnetic properties evaluation is done through Barkhausen noise and permeability measurements.This paper presents a preliminary analysis of the results obtained by all involving laboratories.     

Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level.The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals 2.7 × 10¹² cm⁻² s⁻¹ with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.     

Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.     

Operational feedback on internal structure vibration in 54 French PWRs during 300 fuel cycles

EDF has acquired extensive feedback on vibration of reactor vessel internals by analysing ex-core neutron noise on its 54 pressurized water reactors during the course of over 300 fuel cycles.

This feedback has been built up by processing more than 3,000 vibratory signatures acquired since the startup of its reactors. These signatures are now centralized for the whole of France in the “SINBAD” data base.

Signature processing has enabled:

1. • distinguishing between mechanical phenomena and signature variation linked to unit operation: in particular, the impact on signature level of unit operating parameters such as initial fuel enrichment and burn-up rate was assessed;

2. • among the purely mechanical phenomena, pointing up slight changes in position of vessel internals and the first signs of structural wear: relaxation (in the hold-down spring and fuel rod assemblies) and wear on surfaces of contact between internals and reactor vessel were detected;

3. • lastly and most importantly, automatic recognition of the various types of vibratory behavior of internals.

It was consequently possible to draw up user requirement specifications for automated monitoring of internals, which should soon be integrated in PSAD, a system which groups several reactor monitoring functions.     

钠冷快堆换料系统可靠性研究

由于中子通量以及冷却剂运行温度高,钠冷快中子反应堆（简称钠冷快堆）的换料周期较一般轻水反应堆短。同时,换料过程中隔绝空气的要求以及换料设备本身的复杂性,钠冷快堆只能逐根进行换料,使得总的换料时间较轻水反应堆长。本文采用失效模式与影响分析、故障树分析等方法对典型钠冷快堆换料系统各部分的可靠性进行评价,获得了换料系统每次换料期间的失效概率。基于换料系统各部分失效的影响、失效概率以及恢复时间,分析了换料系统不同失效模式对反应堆运行效率的影响。     

Knowledge-directed characterization of nuclear power plant reactor core parameters

A major life-limiting factor of the UK's Advanced Gas-Cooled Reactors (AGRs) is the condition of the graphite core. Installation of new measurement equipment is difficult and expensive, therefore maximizing the information gained from existing equipment is highly desirable. The main approach to determining the health of an AGR core is through periodic inspections undertaken during planned outages. However, there is the desire to supplement this inspection activity through the analysis of data gathered as part of routine plant operation. One such source of data is measurements taken during refueling and this paper describes knowledge-directed characterization of this refueling data, both spatially across the reactor core and temporally across the operational lifetime of the core. Characterization provides information relating to the current condition of the reactor core and allows suspected ageing trends to be visualized and confirmed. A standard approach for characterizing reactor core data is presented and applied to a variety of different reactor core parameters. The benefit of this approach is that it allows engineers to distill large volumes of refueling data into a readily understandable format in a short period of time. It also allows hypothesized trends relating to the ageing process within the core to be tested and provides supporting evidence for these hypotheses. The trending data is also valuable as it can form the basis of a predictive model of ageing of the reactor core. The ageing process of nuclear graphite is understood from theoretical and experimental viewpoints and this empirical data, gathered from operating reactors, further supports this understanding. This paper represents the initial exploration of using refueling data to construct a predictive model of AGR reactor core ageing.     

Conceptual design of multi-purpose heat reactor “Nuclear Heat Generator”

Small heat reactors can apply to on site demand such as district heat and air conditioning, industrial process heat, greenhouse, and seawater desalination in urban and rural areas. The purpose of this paper is to design conceptually a multi-purpose reactor named “Nuclear Heat Generator (NHG)” which could be installed in energy consuming area. The reactor of 1MWt output is designed without any needs for fuel exchange and decommissioning on site. This cassette typed reactor vessel with sealing is transported to specified fuel fabrication shop every 3 to 4 years in order to exchange used fuels. Steam generators are involved in the self-pressurized integrated reactor with natural circulation. Generated steam pressure from heating reactor is 0.88 MPa (saturated) which is so less than that of current water reactors. Under low steam pressure it is considerably easy to make design of containment vessel and safety device. For economic competition overcoming scale demerit it will be necessary for the cassette type reactor to optimize its system design for the multi-production effect as well as modular construction and recycling system.     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

Use of J–R curves in assessing the fracture behavior of low upper shelf toughness materials

The objective of this investigation was to evaluate the use of small specimen J–R curves in assessing the fracture resistance behavior of reactor vessels containing low upper shelf (LUS) toughness weldments. As required by the U.S. Code of Federal Regulations (10 CFR, Part 50), reactor vessel beltline materials must maintain an upper shelf Charpy V-Notch (CVN) energy of at least 50 ft-lbs (68 J) throughout vessel life. If CVN values from surveillance specimens fall below this value, the utility must demonstrate to the U.S. Nuclear Regulatory Commission (NRC) that the lower values will provide “margins of safety against fracture equivalent to those required by Appendix G of the ASME Boiler and Pressure Vessel Code”. This paper will present recommendations regarding the material fracture resistance aspects of this problem and outline an analysis procedure for demonstrating adequate fracture safety based on CVN values.It is recommended that the deformation formulation of the J-integral be used in the analysis described above. For cases where J-integral fracture toughness testing will be required, the ASTM E1152-87 procedure should be followed, however, data should be taken to 50% to 60% of the specimen remaining ligament. Extension of the crack growth validity limits for J–R curve testing, as described in E1152-87, can be justified on the basis of a “J-controlled crack growth zone” analysis which shows an engineering basis for J-control to 25% to 40% of the specimen remaining ligament. If J-R curve extrapolations are required for the analysis, a simple power law fit to data in the extended validity region should be used. The example analysis performed for low upper shelf weld material, showed required CVN values for a reactor vessel with a 7.8 inch (198 mm) thick wall ranging from 32 ft-lbs (43 J) to 48 ft-lbs (65 J), depending on the magnitude of the thermal stress component.     

Thermal and mechanical characteristics of an instrumented capsule for a material irradiation test

In this study, the thermal and mechanical characteristics are analyzed for the structural integrity evaluation of the instrumented capsule used for the irradiation test of reactor vessel materials in the research reactor, hi-flux advanced neutron application reactor (HANARO). The temperature of test specimens inserted in the capsule mainbody by γ-flux is calculated using a heat transfer code, HEATING 7.2f. The maximum temperature is 556.75 K at the center of the capsule mainbody, thus the temperature satisfies the user's requirement. To estimate the mechanical characteristics of the capsule due to the pressure and thermal loading, stress analysis is carried out with a finite element analysis program, ANSYS. The strength of the capsule's external tube is also evaluated by considering the buckling stress of the capsule mainbody under coolant pressure loading. The results of the analysis show that the temperature distributions are significantly affected by the gap size between the holder and the specimen. The calculated stresses of the capsule structure are well within the allowable stress values of the ASME code. It is expected that the results presented in this paper will be useful in the design and safety evaluation of instrumented capsules for material irradiation tests.     

Coupled thermohydraulic-neutronic instabilities in boiling water nuclear reactors: a review of the state of the art

This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation in an extended operation domain with increased void and thereby increased void reactivity feedback and which often have thinner fuel rods and thereby decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of “unexpected” instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, the subject of BWR instabilities has been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a “new and improved” state of the art has emerged recently.     

VVER-1000型反应堆压力容器热老化分析评估

本文系统介绍了VVER-1000型反应堆压力容器(RPV)的温度监督情况,针对田湾核电站1#机组RPV的温度监督测试结果进行分析,评价运行3年后RPV力学性能(包括拉伸、冲击、断裂韧性)变化行为及热老化脆化机理,评估了当前田湾RPV服役运行后的热老化脆化状态和温度监督的时间安排。结果表明,温度监督样品经过堆内高温环境考验后,焊缝材料表现出一定程度的脆化特征,但母材、热影响区脆化不明显。与康采恩模型的结果和俄罗斯数据相比较后,认为田湾核电站1#机组RPV热老化脆化情况在合理范围内。     

Radiation Service Life of the Vessel Metal in Operating VVER Reactors

The current status of the prediction of radiation embrittlement of the vessel material in first- and second-generation VVER reactors is analyzed. The radiation service life of the vessel of each type of reactor is determined by factors due to the special features of the operating regime of the reactor and the chemical composition of the vessel metal. A method of monitoring the state of the material of first-generation reactor vessels is examined. The method is based on extracting and studying samples of a metal from the inner surface of the sample. The main problems of monitoring the state of the metal in VVER-440/213 and VVER-1000 vessels are analyzed. It is indicated that adjustments must be made in the normative relations which are currently used for predicting radiation embrittlement of vessel material. The most important questions concerning reactor dosimetry for VVER vessel material are illuminated.__________Translated from Atomnaya Energiya, Vol. 98, No. 6, pp. 460–472, June 2005.     

Reactor dosimetry for VVERs RPV lifetime assessment

A state of the art review of Reactor Dosimetry used for reactor pressure vessel irradiation damage assessment and lifetime evaluation of the Russian type VVER reactors is presented. The necessity of prospective studies in Reactor Dosimetry for improvements that will reduce the neutron fluence uncertainty and in this way to substantiate the extension of NPP lifetime is summarized by specialists in Reactor Dosimetry from countries operating VVER reactors such as Russia, Ukraine, Czech Republic, Finland, Hungary, and Bulgaria, together with specialists from Western European countries such as France, Spain, Germany, Belgium, The Netherlands, and UK, operating PWR and BWR type reactors.     

An on-line fuel management method using in-core detector readings for CANDU reactors

An on-line fuel management method for a CANDU reactor has been developed. In the method, the in-core detector readings are used for channel power generation for refueling channel selection. The in-core detector readings are converted to measured mesh readings, and the Kalman filtering technique is applied to reduce calculation and measurement errors of the mesh readings. Then, the estimated channel powers are fed into the refueling channel selection process, in which the channels are refueled so that the difference of zone power from the reference one is minimized. The performance of the method has been demonstrated against the operating data of CANDU 6 reactor. Also, it is found that the core tracking fuel management could be implemented, so that the proposed method would contribute to economic and safe operation of the reactor.     

用Charpy试样研究核压力容器钢的韧性转变温度曲线

用一个标准Ｃｈａｒｐｙ试样及试验过的该试样的两个半截试样所获得的８个“再造Ｃｈａｒｐｙ复合试样”评定核压力容器用Ａ５０８ＣＬ３钢冲击韧性或断裂韧性的温度转变曲线。研究结果表明，用这种方法可得到一条较为可靠而完整的核压力容器钢的冲击韧性或断裂韧性随温度的转变曲线。这相当于在核监测计划中，只需取出一个Ｃｈａｒｐｙ试样即可得到完整的材料性能曲线，这对于提高核压力容器中子辐照脆化监测的可靠性很有价值，特别是对于运行时间较长（堆芯内监测试样数量已经不多）的核容器寿命预测和安全监视尤为重要。     

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

Within the framework of the 6 month WANO program, small samples were cut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime—the ductile-to-brittle transition temperature (DBTT) in the initial state (T_ko) and the phosphorus and copper contents which affect the radiation stability of steel—were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objectives of the study.Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C.The estimated value of T_ko, for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel.The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of T_f obtained from tests before annealing (212 °C) is about 40 °C higher that the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of T_ko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of T_ko = 50 °C.The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.