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.     

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);     

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.     

Vibration-based diagnosis techniques used in nuclear power plants: An overview of experiences

Over the decades, the vibration measurements and the vibration-based diagnosis techniques are widely used in practice and it is now well-accepted tools in various roles to achieve the required performance and the safety of plants. Here author is summarising the experiences on the role and/or the importance of the vibration-based diagnosis in nuclear power plants (NPPs) in a simplified manner which may be useful for engineers and researchers in this research area.     

Finite element methods in the analysis of reactor vessels

The application of the finite element method to linear and non-linear problems in pressure vessel technology is presented. New developments for dealing with components such as liners, prestressing cables and reinforcement are outlined and some improvements possible in thin shell situations are discussed. A general solution technique for non-linear analysis is presented and applied firstly to the problem of the plastic behaviour of steel pressure vessels. The failure of PCRVs by concrete cracking is then considered. Finally, the time-dependent phenomenon of creep is discussed. In all cases the theory is illustrated by practical examples.     

An overview of reactor containment structures

The purpose of this paper is to present an overview of reactor containment structures and to summarize the present state-of-the-art of containment design. The areas covered are types of containments used for nuclear power plants in operation and under construction, and their development. Also presented are codes which currently govern the design, materials, and construction of containments, as well as some thoughts on safety and methods of analysis.     

Structural integrity assessment of aging nuclear reactor pressure vessels

Irradiation embrittlement reduces both the cleavage fracture toughness and the ductile tearing toughness of reactor pressure vessel (RPV) steels. Extensive research programs have investigated the fracture behavior of heavy-section vessels containing flaws. Information obtained from that research has been used to develop regulatory guidance for evaluating the structural integrity of irradiated RPVs. Additional research programs have developed fracture analysis methods, and generated the data required for their implementation. Regulatory guidance employs fracture analysis technology to assure that adequate fracture-prevention margins for RPVs are maintained throughout the licensed operating period of nuclear power plants.     

Matrix displacement methods in fracture mechanics analysis of reactor vessels

The technology of fracture mechanics is developing rapidly in response to increased requirements for integrity of engineering structures. It enables structural engineers to evaluate brittle failure resistance of structures within appropriate regimes of temperature, materials and geometry. The evaluation includes the combined effects of material toughness, flaw characteristics, environment and service loadings. Calculations of stress intensity factors associated with the flaws, geometry and applied loading form the basis of fracture analysis and control procedures for reactor vessels.     

Evaluation of flaws in nuclear piping

This paper summarizes development of flaw evaluation procedures and allowable flaw sizes for piping used in light water reactors, and focuses on the research conducted by the Electric Power Research Institute in support of Section XI of the ASME Code. The paper also presents development of elastic-plastic fracture analysis methods needed in the development of the code evaluation procedures.     

3D analyses of surface flaws in thick-walled reactor pressure-vessels using displacement-hybrid finite element method

Solutions of stress intensity factors for external and internal unpressurized and pressurized surface cracks in internally pressurized thick-walled reactor pressure vessels are determined directly by a three-dimensional displacement-hybrid finite element method. The finite element method is based on a rigorous modified variational principle of the total potential energy, with arbitrary element interior displacements, interelement boundary displacements and element boundary tractions as variables. Special crack front elements, developed using the hybrid displacement model, which contain the proper square root and inverse square root variations of displacements and stresses, are used in this analysis and the three stress intensity factors, K_I, K_II and K_III are solved directly along with the unknown nodal displacements. Stress intensity factor variations for pressurized and unpressurized semi-elliptical inner surface cracks in pressurized cylinders with crack aspect ratios of 0.2 and 1.0, crack depth to cylinder wall thickness ratios of 0.5 and 0.8 and outer to inner diameter ratios of 1.5 and 2.0, are presented. Also, for unpressurized outer surface cracks in pressurized cylinders, the solutions are presented for crack aspect ratios of 0.6 and 1.0, crack depth to cylinder wall thickness ratios of 0.4, 0.6 and 0.8, and outer to inner diameter ratio of 1.5.     

一种用于次临界反应堆中子特性研究的电流灵敏前置放大器的研制

介绍针对用于次临界反应堆中子特性测量的小尺寸(如φ6mm×10mm)3He中子探测器设计的一种基于AD8004芯片的电流灵敏前置放大器的研制,该前放可由长达10m的同轴屏蔽电缆与探头连接,并且能够适应于高计数率情况下中子通量密度的测量,文中给出了具体电路原理图、性能指标参数以及对电路调试的有关问题的讨论。     

Stable crack growth of axial flaws in pressure vessels

The ductile crack growth of axial through and part-through cracks in a vessel under internal pressure has been studied experimentally to contribute to the fundamental problem whether or not and under which conditions resistance curves obtained from specimens can be transferred to large scale components. The experiments and numerical analyses are part of a research program on fracture mechanics failure concepts for the safety assessment of nuclear components.Whereas only an averaged crack extension is determined in specimen tests, the local propagation of cracks may be of main importance for surface cracks in vessels and pipes. In the present experiments, the surface cracks revealed the well known canoe shape, i.e. a larger crack extension has occurred in the axial direction than in the wall thickness direction. Two of these tests have been analysed by finite element calculations to obtain the variation of the J-integral along the crack front and the stress and strain state in the vicinity of the crack. The local crack resistance appeared to depend on the local stress state. To Predict ductile crack extension correctly, J_R-curves have to account for the varying triaxiality of the stress state along the crack front.     

An improved porous media model for nuclear reactor analysis

In this study, two modifications are proposed to mitigate drawbacks of the conventional approach of using the ‘‘Porous Media Model'(PMM) for nuclear reactor analysis. In the conventional approach, whole reactor core simplifies to a single porous medium and also, the resistance coefficients that are essential to using this model are constant values. These conditions impose significant errors and restrict the applications of the model for many cases,including accident analysis. In this article, the procedures for calculating the coefficients are modified by introducing a practical algorithm. Using this algorithm will result in obtaining each coefficient as a function of mass flow rate.Furthermore, the method of applying these coefficients to the reactor core is modified by dividing the core into several porous media instead of one. In this method, each porous medium comprises a single fuel assembly. PMM with these two modifications is termed ‘‘multi-region PMM' in this study. Then, the multi-region PMM is introduced to a new CFD-based thermo-hydraulic code that is specifically devised for combining with neutronic codes.The CITVAP code, which solves multi-group diffusion equations, is the selected as the neutronic part for this study. The resulting coupled code is used for simulation of natural circulation in a MTR. A new semi-analytic method,based on steady-state CFD analysis is developed to verify the results of this case. Results demonstrate considerable improvement, compared to the conventional approach.     

Status report on intergrated analysis methods for nuclear reactor structural analysis

The purpose of this paper is to look at the analysis of a nuclear reactor system as a total analysis task and to examine the assumptions which are made in the separation of the analysis task into its component disciplines. The structural analysis discipline is then examined in more detail to try to define a workable approach to an integrated structural analysis of the reactor system. We will start with a general discussion of the total analysis task, starting from the initial concept of the reactor plant. The total task will then be subdivided into the respective disciplines and an attempt will be made to rationalize or criticize the division into separate disciplines. The discipline of structural mechanics will then be examined in view of its interactions with other disciplines such as fluid flow and nuclear analysis to determine the degree of coupling which exists among these disciplines. This will be done by examining the interactions of the state variables which apply at each point of the system. The state variables considered will include fluence, temperature, displacement and pressure. The state variables defined will then be used as the basis for the definition of an overall structural model of the reactor system. Such an overall model can be conceived in terms of the present status of analytical techniques, by the use of such concepts as substructuring, constraint equations, coupled solutions for heat transfer, stress and dynamic analysis, along with fourth generation computer capabilities. A block design for an overall structural model will be discussed and also the areas which require new analysis techniques. The last section will present an outline of a mode of operation of a structural design/analysis activity which is established to implement a comprehensive integrated structural analysis of an entire reactor system. The concept of an evolving model of the system will be presented and the coordination required to successfully manage such a design/analysis approach will be discussed. A brief discussion of the effects of non-linear effects such as creep, plasticity, gaps on the overall approach will be included.     

An international scale for assessing the danger of nuclear plant events

State Nuclear Power Supervisory Committee Scientific and Technical Center for Industrial and Nuclear Power. Translated from Atomnaya Énergiya, Vol. 70, No. 1, pp. 3–8, January, 1991.     

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.     

The dynomak: An advanced spheromak reactor concept with imposed-dynamo current drive and next-generation nuclear power technologies

A high-β spheromak reactor concept has been formulated with an estimated overnight capital cost that is competitive with conventional power sources. This reactor concept utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt (FLiBe) blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER-developed cryogenic pumping systems were implemented in this concept from the basis of technological feasibility. A tritium breeding ratio (TBR) of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%.