Influence of pre/in-service inspections and tests on the reliability of reactor pressure vessels

The influence of the following actions on the probability of brittle failure of the reactor pressure vessels will be estimated by probabilistic fracture mechanics: ultrasonic inspection of the welds; hydro-test of the vessel; and crack growth by normal, upset and test conditions. Taking into account that the in-service inspections and tests are done at short intervals the reliability can be shown to be extremely high.     

Boat sampling and inservice inspections of the reactor pressure vessel weld No. 4 at Kozloduy NPP, Unit 1

The paper deals with reactor pressure vessel (RPV) boat sampling performed at Kozloduy Nuclear Power Plant, Unit 1, from August to November 1996. Kozloduy NPP, Unit 1 has no reactor vessel material surveillance program. Changes in the material fracture toughness resulting from the fast neutron irradiation which cannot be monitored without removal of the vessel material. Therefore, the main objective of the project was to cut samples from the RPV wall in order to obtain samples of the RPV material for further structural analyses. The most critical area, i.e. weld No. 4 was determined as a location for boat sampling. Replication technique was applied in order to obtain precise determination of the weld geometry necessary for positioning of the cutting tool prior to boat sampling, and determination of divot depth left after boat sampling and grinding of sample sites. Boat sampling was performed by electrical discharge machining (EDM). Grinding of sample sites was implemented to minimize stress concentration effects on sample sites, to eliminate surface irregularities resulting from EDM process, and to eliminate recast layer on the surface of the EDM cut. Ultrasonic, liquid penetrant, magnetic particles, and visual examinations were performed after grinding to establish baseline data in the boat sampling area. The project preparation activities, apart from EDM process, and the site organization lead was entrusted to INETEC. The activities were funded by the PHARE program of the European Commission.     

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 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.     

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.     

Recent developments in the finite element analysis of prestressed concrete reactor vessels

This paper describes recent developments in the nonlinear deformation and ultimate load analysis of prestressed concrete reactor vessels using finite elements. First, a number of finite element models are called into attention for the idealization of composite structures such as reinforced and prestressed concrete components. Then different inelastic constitutive models are proposed for the behaviour of concrete in the pre- and post-failure regime. Subsequently various numerical techniques are examined for the solution of nonlinear problems, especially with regard to their distortion of the constitutive model. In conclusion these modelling techniques are applied to the analysis of four typical examples, the nonlinear deformation analysis of a concrete specimen subjected to biaxial compression, the crack analysis of a thick-walled concrete cylinder, the overload analysis of the THTR 1 : 5 scale model, and the ultimate load analysis of a concrete top closure model.     

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.     

Material properties for reactor pressure vessels and containment shells under dynamic loading

The effects of high strain rate, dynamic biaxial loading and deformation mode (tension, shear) on the mechanical properties of AISI 316 austenitic stainless steel in as-received and pre-damaged (creep, LCF) conditions are reported. This research was conducted to assess the performances of the containment shell of fast breeder reactors. The results of this research have been utilized to prepare similar investigations for SA 537 Class 1 ferritic steel used for the containment shell of LWR. The first results of these investigations are reported. A programme to study the mechanical properties of plain concrete with real size aggregate at high strain rate is described.     

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.     

Prediction of lower bound fracture toughness for lifetime evaluation of reactor pressure vessels

The potential damage of embrittlement in service is a very important problem of MnMoNi steels used for the nuclear reactor pressure vessel. A decrease of critical flaw size may occur when embrittlement proceeds. The remaining lifetime of the reactors should be assessed taking into account the embrittlement of the steel paying special attention to the degradation of dynamic fracture toughness. The present study introduces the basic concept of the remaining lifetime assessment. Examined was a small specimen fracture toughness test for measuring the dynamic fracture toughness of nuclear reactor pressure vessel (RPV) steels. The result was applied in the measurement of the dynamic fracture toughness of 12 heats of RPV steels. The test results were analyzed to find more practical applications and a method is presented to predict the lower bound dynamic fracture toughness using the Charpy impact test and tensile test results.     

Design and analysis of concrete reactor vessels: New developments, problems and trends

This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance — areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of these problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep.     

Applications of probabilistic fracture mechanics to light water reactor pressure vessels and piping

This paper reviews recent calculations of the statistical reliability of LWR reactor vessels and piping. The broad theoretical principles of these calculations are well established and it is therefore possible to compare the physical assumptions made in different calculations. Such a comparison shows that certain functions are not known at all well; for example, (i) the frequency of occurrence of cracks in weld-regions, (ii) the size distribution of cracks, (iii) the efficiency of methods of non-destructive examination and (iv) the transient loadings that the system experiences in service. On the other hand, relevant materials properties (toughness, crack growth characteristics) appear to be known adequately if not completely. Despite these quantitative uncertainties in the input, it seems possible to draw several broad conclusions from the results of these calculations. These concern (i) the low absolute rates of failure, (ii) the way these depend upon time in service, (iii) the effect upon them of in-service inspection and (iv) their sensitivity or otherwise to the physical assumptions which are made.     

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.     

Manufacturing of ultra-large diameter 20 MnMoNi 5 5 steel forgings for reactor pressure vessels and their properties

As the structural material for RPV typical of increased dimensions, a set of ultra-large diameter steel forgings for a PHWR RPV is presented as outlined below.

1. (1) Material designation: 20 MnMoNi 5 5 (similar to SA508, Cl.3)

2. (2) Size of the forgings: flanges, 8.440 mm OD, a weight of 238 tons for shell flange; shells and torus, 7,920 mm OD, with large height; cover dome, 6,800 mm OD in chord and 460 mm thick; blank before formed to dome is ca. 8,000 mm OD.

3. (3) Chemical composition: particular effort was made for minimizing the tramp elements as P, S, As, Sn, Sb, Cu.

4. (4) Manufacturing, key points: steel making - combined refining and degassing in ladle; ingot making - largest size ingots, including 570 ton and 500 ton ingots; forging - special “outside-the-press” forging and forming techniques; heat treatment - prevention of H₂ flaking in normalizing and tempering and handling of the extra-large forgings at water quenching.

5. (5) Metallurgical properties: sufficiently uniform carbon distributions in the forgings; a lowest possible content of hydrogen, non-metallic inclusions and oxygen.

Mechanical properties: uniformity in tensile and toughness properties; flaws - only limited number of spots of UT indications under 2 mm EFG (EFS).     

Seismic analysis of heavy-liquid-metal-cooled reactor vessels

The linear-elastic seismic analysis for the vessel of a heavy-liquid-metal reactor was undertaken based on the Design Response Spectrum (DRS) approach and for a 0.5 g earthquake. Four support types for the vessel were analyzed and it was found that the roll support exhibits the best overall performance. The variation of the normal-mode frequency and total peak stress intensity with the vessel diameter and thickness was also studied. It was found that the frequency of the first normal mode increases with increasing vessel diameter and thickness, while the total peak stress intensity decreases with increasing vessel thickness and is roughly independent of the vessel diameter. Two new dimensionless groups are introduced that enable correlation of the frequency and stress-intensity data with adequate accuracy. It is proposed that these correlations be used for quick estimate of the seismic response of a vessel of arbitrary size, material and contained (heavy) fluid, subject to an earthquake of arbitrary magnitude.     

Seismic analysis of cracked reactor vessels

This paper deals with the dynamic response of a thin finite, elastic circular cylindrical shell representing a reactor vessel to time-dependent loadings symmetrical with respect to the axis of the cylinder. The shell contains an axial through-crack of length 2c. The dynamic counterpart of Donnell's shell equations are used in this investigation. Extensive numerical results are presented for stress intensity factors in aluminum and steel vessels and results are discussed.     

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.