LWR Plants Research Articles

Finite Element Analysis of Internally Circumferential Surface Cracked Pipe Stainless Steel under Bending and Torsion Loadings

Early Light Water Reactor Plants (LWR plants) in Japan are nearly 30 years old and utilities have been conducting technical evaluations of facility aging. Accurate prediction of fracture behavior for ductile materials is one of the most important issues to guarantee the integrity of the structure such as pipes in plants. If a crack is detected during the in-service inspections, the structural rational maintenance of the cracked pipe has been promoted based on the defect evaluation methods specified ASME Boiler and Pressure Vessel Code and JSME Rules on Fitness-for-Service for Nuclear Power Plants. Recently, experiments and simulations have been performed on pipes subjected to bending and torsion loadings. However, the failure criterion is still under investigation, and it is significant to propose a method to evaluate it appropriately. In this study, a finite element model of a large circumferentially cracked stainless steel pipe was generated and elastic-plastic analysis of stationary crack and crack extension were performed to validity of the results.

Review of Notch Effect on Fatigue Behavior of Materials for LWR Plants in High Temperature High Pressure Water

Review of Notch Effect on Fatigue Behavior of Materials for LWR Plants in High Temperature High Pressure Water

Time for a game-changing nuclear technology

Time for a game-changing nuclear technology

Converting Maturing Nuclear Sites to Integrated Power Production Islands

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on‐site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage of already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on‐site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.

CORIUM COOLABILITY UNDER EX-VESSEL ACCIDENT CONDITIONS FOR LWRs

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+ LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

Water Chemistry of LWR Plants (9)

Water Chemistry of LWR Plants (9)

Water Chemistry of LWR Plants (7)

Water Chemistry of LWR Plants (7)

ICONE15-10472 COOLABILITY ANALYSIS OF BOTTOM-FED DEBRIS BEDS IN A LWR SEVERE ACCIDENT

This paper investigates the potential effectiveness of natural circulation-driven coolability (NCDC) as a severe accident mitigative measure. The NCDC can particularly be useful in LWR plants which employ external cavity flooding. The main idea is to provide a simple design solution that facilitates bottom feeding of coolant into the debris bed, and uses steam production in the decay-heated debris bed to drive the two-phase flow natural circulation. We use an analytical one-dimensional model to calculate characteristics of two-phase thermal-hydraulics in porous media. The model employs Lockhart-Martinelli correlations for two-phase flow friction and void fraction, and Ergun's correlation used for single-phase flow resistance. Adaptation and verification of the model are discussed in the paper. Coolability of debris beds with coolant bottom-fed is evaluated for a broad range of conditions. The analysis suggests that the dryout heat flux (DHF) in bottom-fed configurations can be increased by 80% to 160%, when compared to DHF in top-flooding beds.

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4x 10−2 (mean) and 3.9x 10−2 (median) for the BWR suppression pool case, 2.2x10−3 (mean) and 2.8x10−10 (median) for the BWR pedestal case, and 6.8X10−2 (mean) and 1.4x10−2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

The European utility requirement document (EUR)

The major European electricity producers work on a common requirement document for future LWR plants since 1992. They aim at requirements acceptable together by the owners, the public and the authorities. Thus the designers can develop standard LWR designs acceptable everywhere in Europe and the utilities can open their consultations to vendors on common bases. Such a standardisation promotes an improvement of generation costs and of safety; public and authorities acceptance should be improved as well; significant savings are expected in development and construction costs. Since the early stages of the project, the European utility requirement (EUR) group has grown significantly. It now includes utilities from nine European countries. Utilities from two other European countries are joining the group. Specific cooperation agreements are also in progress with a few extra-European partners.

Current topics of the structural design for LWR plants in Japan

The major movements related to the structural design of light water reactors in Japan are the system change of the structural standards along with the deregulation of the electric utility industry and the evaluation of plant life management. This paper summarizes these situations, followed by a detailed description of the research and development of technologies and the development of voluntary standards for each. It is considered that the results of the research should be incorporated in the voluntary consensus standards.

Plant life management activities of LWR plants in Japan

Plant life management activities of Japanese LWR plants have been conducted since the early 1990s by the utilities and MITI (Ministry of International Trade and Industry) cooperatively. In Japan, where the regulatory practices are different from those in the US, there is neither law nor regulation that prescribes a licensed plant life for nuclear power plants. When an annual inspection is completed without any problem, the next cycle of operation would be permitted and this cycle can be repeated. However, it is generally known that mechanical components and structures deteriorate as they get older. So, we consider it very important to evaluate the long-term integrity of major systems, structures and components of old nuclear power plants. Japanese plant life management study consists of two parts. Both parts of the study were carried out confirming the integrity for the long-term operation of the three oldest Japanese LWR plants: Tsuruga Power Station Unit No.1 (BWR), Mihama Power Station Unit No.1 (PWR) and Fukushima Dai-ichi Nuclear Power Station Unit No.1 (BWR). The Part 1 study was conducted for the purpose of obtaining an outlook for long-term safety operation and was completed in 1996. The Part 2 study was conducted ensuring the plant integrity for the long-term operation in terms of, not only safety, but also reliability. The results of the Part 2 study were made public in February, 1999. Then, the recommended maintenance items were to be added to the existing maintenance programs of the three LWR plants.

Developments in gamma scanning irradiated nuclear fuel

Non-destructive methods, based on high-resolution gamma-ray spectroscopy, have been further developed for on-site measurements of several nuclear fuel parameters and efficient verification of the fuel in-pile performance. The measuring system is based on a high-purity germanium detector (HPGe) together with suitable fast electronics and an on-line PC data aquisition module. Measurements have been carried out in the fuel storage pools of several LWR plants by the ABB Atom Company for more than 10 years. Special computer codes for evaluation of the spectroscopic data and detailed comparison with core physics evaluations have been developed. Techniques have been employed for determination of fission gas release (FGR), showing good correlation with destructive post-irradiation examinations (PIE) using mass spectrometry. Axial cesium redistribution which gives a control on local temperature history (possibly overheating), axial burnup profiles, burnup comparison between fuel rods and examination of the fuel rods with respect to their integrity are other applications. Relative axial power distribution, using gamma radiation from 140La, can be determined with a precision of ±1% for a single rod node measurement. It is concluded that this high-resolution gamma scanning technique that has been developed is a very versatile measuring method. It rapidly gives information about several physical processes in the fuel, with the possibility of feedback to reactor operators.

A review of fatigue failures in LWR plants in Japan: Iida, K. Nucl. Eng. Des. (Dec. 1992) 138 (3), 297–312

A review of fatigue failures in LWR plants in Japan: Iida, K. Nucl. Eng. Des. (Dec. 1992) 138 (3), 297–312

A review of fatigue failures in LWR plants in Japan

A review was made of fatigue failures of nuclear power plant components in Japan, which were experienced in service and during periodical inspection. No case has been recently reported of a service fatigue failure of a reactor pressure vessel itself, excluding nozzle corner cracks, that occurred many years ago. But, service fatigue failures have been occasionally experienced in piping systems, pumps, and valves, on which fatigue design seems to have been inadequately applied. The causes of fatigue failures can be divided into two categories: mechanical-vibration-induced fatigue and thermal-fluctuation-induced fatigue. Vibration-induced fatigue failure occurs more frequently than is generally thought. The lesson gleaned from the present survey is a recognition that a service fatigue failure may occur due to any one or a combination of the following factors: (1) lack of communication between designers and fabrication engineers, (2) lack of knowledge about a possibility of fatigue failure and poor consideration about the effects of residual stresses, (3) lack of consideration on possible vibration in the design and fabrication stages, and (4) lack of fusion or poor penetration in a welded joint.

Sensitivity Analysis on Safety Limits of Large FBR Cores for Seismic Vertical Vibration of Control Rods

Liquid metal fast breeder reactor (LMFBR) plants are necessary to have system components with thin wall structures that are much different from those of LWR plants, and so some reactivity vibrations are induced by seismic vertical movements between the core and the control rod system. Dynamic response analyses during these vibratory transients were performed and the maximum allowable displacements of the core relative to the control rod system were evaluated with a view to avoid fuel failures. Two two-region homogeneous and axially heterogeneous oxide cores, and a two-region homogeneous metal core were taken as the reference large FBR cores. The amplitude and the frequency of control rod vibrations, three kinds of reactivity coefficients and scram conditions were selected as main parameters for sensitivity analyses, and integrity limits of fuel assemblies were evaluated and the results of reactivity amplitude were converted into vertical displacements to have the target requirements for the aseismic design. The allowable maximum reactivity amplitudes were 1.19, 1.15 and 0.89$, and the corresponding allowable relative displacements between the core and the control rod system were 44, 52 and 42 mm, respectively for the above-mentioned three cores in the severest condition. Effects of seismic vibration characteristic, control rod scram time and other affecting factors were discussed based on the final results.

Sensitivity Analysis on Safety Limits of Large FBR Cores for Seismic Vertical Vibration of Control Rods.

Liquid metal fast breeder reactor (LMFBR) plants are necessary to have system components with thin wall structures that are much different from those of LWR plants, and so some reactivity vibrations are induced by seismic vertical movements between the core and the control rod system. Dynamic response analyses during these vibratory transients were performed and the maximum allowable displacements of the core relative to the control rod system were evaluated with a view to avoid fuel failures. Two two-region homogeneous and axially heterogeneous oxide cores, and a two-region homogeneous metal core were taken as the reference large FBR cores. The amplitude and the frequency of control rod vibrations, three kinds of reactivity coefficients and scram conditions were selected as main parameters for sensitivity analyses, and integrity limits of fuel assemblies were evaluated and the results of reactivity amplitude were converted into vertical displacements to have the target requirements for the aseismic design. The allowable maximum reactivity amplitudes were 1.19, 1.15 and 0.89 $, and the corresponding allowable relative displacements between the core and the control rod system were 44, 52 and 42 mm, respectively for the above-mentioned three cores in the severest condition. Effects of seismic vibration characteristic, control rod scram time and other affecting factors were discussed based on the final results.

Ductile fracture analysis of carbon steel pipe with a circumferential through-wall crack

It is necessary to make clear the pipe fracture conditions based on elastic-plastic fracture mechanics to assess the leak before break situation of carbon steel pipes for LWR plants. The aim of the present work is to discuss the effects of pipe size, initial crack length and fracture toughness on the estimated fracture load and mode of carbon steel pipes with a circumferential through-wall crack. As an analytical method, the R6-Rev.3 approach was applied to the pipe fracture analyses considering its simplicity in the application. The results indicate that the net-section collapse attainment becomes difficult with increasing diameter and decreasing thickness of carbon steel pipes. The degree of the net-section collapse attainment decreases with increasing crack length up to some critical size and then increases. The predicted fracture load is more sensitive to the material's J– R curve than to the elastic-plastic fracture toughness J IC . And a simple limit load analysis based on the yield stress is appropriate to evaluate the fracture load, as long as a proper margin was included.

Variation in Fracture Toughness of Carbon Steel due to Test Standards and Its Influence on Fracture Load Prediction.

This paper describes an apparent difference in the fracture toughnesses obtaind by using JSME S 001 and ASTM E 813 test standards, and its influence on a fracture load prediction of carbon steel pipes. Fracture tougnhness tests were conducted in air at room temperature on 1CT specimens prepared from a carbon steel pipe STS 42 (20 B, sch. 100) for LWR plants. Using these fracture toughnesses, analyses were carried out to evaluate the fracture loads of a carbon steel pipe with a circumferential through-wall crack. It is noted that the predicted fracture loads for the two pipes with same geometry are almost the same in spite of large difference in the fracture toughnesses.