Annealing of nuclear reactor pressure vessels

The neutron embrittlement that occurs in the beltline of reactor pressure vessels (RPV) can be managed by various techniques such as fuel management, but only thermal annealing can reverse the effects and result in a restoration of RPV beltline material toughness. The US Nuclear Regulatory Commission has recently revised the Code of Federal Regulations to include the use of thermal annealing of RPV for recovery of material toughness. The Annealing Rule, 10 CFR Part 50.66, has an associated Regulatory Guide 1.162 that describes the format and content of a thermal anneal report that must be submitted to the NRC prior to performing an anneal. This paper will describe the thermal annealing process including regulatory requirements in 10 CFR Part 50.66, techniques for predicting and measuring the toughness recovery, and NDE requirements. Although 14 Russian-designed RPVs have been annealed, there are sufficient differences between the Russian and US designs to question the ease of thermal annealing without producing any unwanted dimensional changes in the RPV and associated piping. The paper will discuss the ongoing annealing demonstration project supported by the Department of Energy which performed a thermal anneal on a canceled pressured water reactor at Marble Hill, Indiana. The associated NRC programs also will be described. This annealing demonstration will be used to bench mark the expected thermal and stress distributions created by thermal annealing and minimize the possible dimensional changes in the RPVs. The paper also will discuss the first possible implementation of thermal annealing for a US commercial nuclear power plant and some important issues that will need to be addressed.     

Structural integrity assessment of the reactor pressure vessel under the external reactor vessel cooling condition

Since the suggestion of external reactor vessel cooling (ERVC), the effects of melting and cooling on the response of structural integrity of the reactor pressure vessel (RPV) under core melting accident conditions have been investigated. To investigate the initial behavior of RPV lower head and the effects of analysis conditions on the structural integrity of RPV, the transient analysis is utilized considering the transient state. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The temperature and stress analyses are performed for core melting accident by using ABAQUS. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. In addition, the results by transient analysis are compared with those by steady state analysis and the effects of analysis conditions on structural integrity are reviewed.     

Structural engineering of pre-stressed reactor pressure vessels

Following a historical survey of the development of pre-stressed concrete reactor pressure vessels, a review of major engineering design and analysis problems is given.     

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of . Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.     

Proof pressure tests of the first reactor vessels at Hartlepool and Heysham nuclear power stations

The nuclear stations currently nearing completion at Hartlepool and Heysham are the world's first design of Advanced Gas Cooled Reactor to use podded boiler construction. The proof pressure tests on the first reactor pressure vessels at both stations were carried out during the early part of 1980. The object of the proof pressure test is to satisfy statutory and contractual requirements by demonstrating the integrity of the completed concrete pressure vessel, liners and pressurised penetrations when subjected to a pneumatic test pressure of 740 lbf/in² g, equivalent to 1.15 times the design pressure of 644 lbf/in² g. Testing was carried out in the latter period of the construction phase of each station with the majority of the plant and buildings in an advanced state of construction and the reactor internals essentially complete. This paper examines the results of both reactor 1 vessel tests and compares the behaviour of the vessels with one another and with analytical predictions. Possible sources of instrumentation error are identified and discussed. The conclusions are drawn from correlation of the test results with theoretical predictions.The vessel pressure was raised and lowered in six stages to and from the test pressure with a hold period between each stage to read instruments, examine the vessel surfaces and leak check the penetrations and closures. To obtain the relevant information at each pressure hold, the vessels' permanent instrumentation, consisting of vibrating wire strain gauges and their associated electrical resistance thermometers, and the liner and reactor internal thermocouples were monitored. In addition, strain gauges fixed to the concrete face of the liner during construction were monitored and liner strains measured during the test were compared with predictions. Vessel deflections were monitored by special equipment mounted on the reactor building to enable vessel dilation to be measured. The vessel top and bottom caps were instrumented by manometric systems in order that the cap profiles and deflections could also be determined.The measurements and observations made during the proof pressure tests adequately demonstrated that the vessels behaved in accordance with predictions. The concrete strains and vessel external deflections confirmed that the vessels behaved in a linear and elastic manner throughout the tests, with no cracking being observed in the concrete during the tests. Liner strains compared favourably with predicted values, exhibiting a linear behaviour under increasing pressure. The strain levels recorded gave complete confidence in the liner design.The tests confirmed the integrity of the vessels, thus enabling them to go forward to the engineering run stage in the commissioning programme leading ultimately to the raising of power.     

Thermal and mechanical stresses in nuclear reactor vessels

This paper is concerned with stress and deformation analysis of a circular cylindrical, thin, elastic shell, representing a nuclear reactor vessel, with an insulated cutout of arbitrary shape subjected to mechanical and thermal loads. The analysis is based on the method of superposition. The actual stresses in the shell may be considered as the sum of the following two parts:

1.

(a) Stresses caused by the given heat flow and mechanical loads in a similar shell without a hole (nominal stress solution);     

A review of the ASME low upper shelf toughness evaluation procedures for nuclear reactor pressure vessels

The ASME Section X1 Working Group on Flaw Evaluation has proposed criteria for the evaluation of reactor pressure vessel beltline materials which have an upper shelf energy less than 50 ft-lbs (69 J). These criteria have been assessed and applied to Linde 80 weld materials in recent investigations; this assessment and evaluation are described in the paper.

A key element in the evaluation procedure is the J_R curve for the relevant material. Recent experimental studies have demonstrated that the J_R curve is size dependent for some materials, in the sense that the J_R curve slope decreases with increasing specimen thickness. This paper assesses this experimental work and discusses it in the context of the integrity of nuclear reactor pressure vessels.     

Experience of the integrity assessment of the highly embrittled reactor pressure vessel for the life extension

The reactor pressure vessel (RPV) is the key component of pressurized water reactor. It has to comply with various rules and regulatory guides to ensure sufficient safety and operating margins during the plant lifetime. Thus, it is crucial to assure the integrity of RPV for an effective plant lifetime management program. In this paper, the status and the experiences of various integrity issues of the highly embrittled RPV are introduced. A circumferential weld in the beltline region of the Kori Unit 1 RPV was projected to be unable to satisfy the minimum upper-shelf energy requirement and the reference temperature-pressurized thermal shock requirement before the end of 40-year design lifetime. The detailed integrity assessments had been performed to resolve both issues and the results summarized. In addition several actions have been taken as aging management programs to assure the integrity of Kori Unit 1 RPV during the extended operation. Details of the activities such as, redefining initial reference temperature-nil ductility transition temperature, installing ex-vessel dosimetry, and withdrawal and testing of additional surveillance capsule are explained. Finally, the applicability of these and other activities including thermal annealing to mitigate the effects of the irradiation embrittlement are evaluated.     

Structural integrity assessment of pressure tubes for Wolsong Unit 1 based on operational experiences

Pressure tube integrity has been considered as a key issue since the first operation of the CANDU reactor. Wolsong Unit 1 has been in service since 1983 and subjected to inspections three times covering 44 tubes. The in-service inspections revealed that a major portion of inspected tubes was in contact with calandria tubes. This is likely to increase the probability of blister formation which is a potential threat to pressure tube integrity. Fortunately, the inspection results indicated that no tube has been affected by blister formation so far. Nevertheless, to reduce the undue risk of blister formation the utility has decided to conduct spacer relocation every year until the entire core is covered. On the other hand, LBB analysis of pressure tubes using AGS performance measured at Wolsong Unit 2 indicated that the operational safety margin was marginal when using 15-year operational data. This raises the concern of pressure tube integrity at Wolsong Unit 1 which has a more than 15-year operation. This paper presents the overall integrity evaluation of Wolsong Unit 1 pressure tubes considering AGS performance test results and operating experience data.     

Radiation-induced embrittlement in light water reactor pressure vessels

In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integrity is a significant economic consideration because the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant. In addition to plant life considerations, LWR-PV embrittlement creates significant cycle-to-cycle impact through the restriction of normal heat-up and cool-down reactor operations.Recent LWR-PV benchmark experiments are analyzed. On this basis, it is established that an exponential representation accurately describes the spatial dependence of neutron exposure in LWR-PV. Implications produced by this simple exponential behavior are explained and trend-curve models for the prediction of PV embrittlement are derived. These derivations provide for a clearer understanding and assessment of the assumptions underlying these trend-curve models. It is demonstrated that LWR-PV embrittlement possesses significant material dependence.     

Apparative developments for inservice inspections of reactor pressure vessels

Inservice inspections of primary circuit components are important preventive measures to guarantee nuclear power plant integrity, satisfying simultaneously reactor safety and economy in plant operation. Emphasizing pressurized water reactor pressure vessel (RPV) inspections, recent developments of new generations of automated and mechanized ultrasonic inspection equipment are presented. Starting from general equipment design and inservice implementation criteria, specific examples are given. Main attention is directed to equipment realization of phased array and ALOK inspection techniques, especially in their combination. Refined aspects of subsequent computer processing and evaluation of defect detection data are described. Analytical features and potential for further developments become evident. Remote controlled RPV inspections are stressed by describing a new generation of central mast manipulators, forming an integral part of total inservice inspection system.     

The integrity of the pressure containment of the high-temperature reactor

The safety concept for ensuring the integrity of the pressure retaining containment is determined by the structural and system-specific inherent safety characteristics and features of the high-temperature reactor. The integrity of the pressure retaining containment, i.e. the elimination of a major failure, is achieved by a system of measures ensuring a high standard of quality and safety. The fundamental cornerstones of this safety concepts are the stringent requirements in the design and manufacture in view of an optimized production technology as well as specific structural solutions such as, e.g., the prestressed concrete reactor vessel. Additional safety measures such as the quality control performed independently of the manufacturer's works and the in-service inspection, have to be considered as redundant safety measures. The in-service inspection can be limited to the confirmation of safety-relevant data and analysis of deviations from these data. Recurrent non-destructive tests within the PCRV are not required, however, possible to a certain extent.     

Interpretation of strain measurements on nuclear pressure vessels

Selected results from strain measurements on four nuclear pressure vessels are presented and discussed.The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts.The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as a detailed knowledge of the behaviour of the signal from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions.     

Experience of irradiation embrittlement in the Loviisa reactor pressure vessels

The water gap between the wall and the core of the RPV (Reactor Pressure Vessel) in a VVER-440 plant is small compared with typical Western type LWR₅. The neutron fluence on the RPV wall is, consequently, much higher in a VVER-440 plant. In older VVER-440 plants the material of the RPV, especially the horizontal core weld, contains so much impurities (P- and Cu-content) that the irradiation embrittlement has become a problem. On bases of fracture mechanics analyses in Loviisa, IVO has been forced to make several measures to ensure safe operation of the plants. According to IVO's current understanding, both plants may be in operation for the design life without annealing of the RPVs.     

Structural homogenized analysis for a nuclear reactor core

A nuclear reactor core is composed of a great number of tubular beams with periodic structure, which are immersed in an acoustic fluid. In the present paper, a 3-D homogenization model is developed to predict its overall dynamic behavior. An approximate solution to the local problem is given. The application to an 1-D example shows that approximate expressions of the natural frequency, the added fluid mass and the equivalent sound speed can be used in engineering estimation.     

Study on microstructural changes in thermally-aged stainless steel weld-overlay cladding of nuclear reactor pressure vessels by atom probe tomography

The effect of thermal aging on microstructural changes was investigated in stainless steel weld-overlay cladding composed of 90% austenite and 10% δ-ferrite phases using atom probe tomography (APT). In as-received materials subjected to cooling process after post-welding heat treatments (PWHT), a slight fluctuation of the Cr concentration was already observed due to spinodal decomposition in the ferrite phase but not in the austenitic phase. Thermal aging at 400 °C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the ferrite phase. The chemical compositions of M₂₃C₆ type carbides seemed to be formed at the austenite/ferrite interface were analyzed. The analyses of the magnitude of the spinodal decomposition and the hardness implied that the spinodal decomposition was the main cause of the hardening.     

严重事故条件下压力容器完整性评价的研究进展

堆芯熔融物堆内滞留(In-Vessel Retention,IVR)是以AP1000为代表的第三代轻水反应堆严重事故管理的重要策略之一,也是严重事故条件下保证压力容器完整性(Reactor Vessel Integrity,RVI)的典型方法之一.该文综述了国外在严重事故条件下压力容器完整性试验研究和理论分析的现状,总...     

Research on structural integrity of pressure boundary in light water reactor designs

This paper is a review of the recent researches performed and planned in Japan relevant to the structural integrity of the pressure boundary in light water reactor designs. Various aspects of relevant work on materials, pressure vessel and piping models are described.