A multidecade linear reactor simulator

Towards an elastic-plastic fracture mechanics predictive capability for reactor piping

The Reactor Management System (RMS), an on-line system with a minicomputer, has been in operation at a boiling water reactor (BWR) nuclear power plant since 1977. The objectives of this system are to perform detailed monitoring and prediction of the core status and to make reactor operation more efficient, simpler, and easier. One of the features of the system is that the RMS utilizes process computer calculated data (such as power distribution and exposure distribution in the core) transferred through the data link, as well as the plant data (such as local power range monitor readings and control rod positions) transferred through the process input/output system. Based on operational experience at the BWR plant, calculation models have been improved to achieve higher accuracy, and new functions have been added to fulfill the operator's demands. The system has now become a useful tool for the operator. For instance, the power level prediction function has become essential for plant operation at the control rod pattern change.

Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material aging issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation, based on the appropriate recognition of the neutron irradiation embrittlement, is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complemental use of the surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old boiling water reactor (BWR) RPV steels with high impurity content. Some discrepancies of pressurized water reactor (PWR) RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201-2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long term surveillance program,” which is to be applied for the operation beyond 40 years. The use of the reloading, reirradiation, and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201-2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan.

A series of tests to establish the feasibility of monitoring relief valve positions using acoustic sensors is described. The sensors respond to noise generated when steam flows through the valve. Six types of safety and relief valves commonly used in light water reactor plants were instrumented and actuated under simulated plant conditions. In addition, background noise was measured in both pressurized water reactor plant and a boiling water reactor plant so that relative signal levels (valve open versus valve shut) could be determined. Results show that acoustic levels with the relief valves open are from one hundred to one thousand times greater than the levels with the valves shut, depending upon the valve type. Data are presented that will help system designers predict the acoustic levels that will be encountered when a relief valve opens.

This paper describes the validity of successive inspections in a chain of the integrity assessment applying Fitness-For-Service (FFS) Codes to piping with stress corrosion cracks. Integrity assessment of primary recirculation system (PLR) piping with stress corrosion cracks is one of the recent interests in Japanese boiling water reactor (BWR) plants. Girth weld joints with detected stress corrosion cracks are required to secure the integrity during subsequent 5 years operation and to conduct three times of successive inspections for PLR piping. If the safety factor included in FFS Codes works effectively, this, like frequent successive inspections are deemed unnecessary. However, since there is some uncertainty in the measurement of crack size and in crack growth estimation in the integrity assessment of piping with stress corrosion cracks, a possibility of failure cannot be denied when not conducting successive inspections to confirm the validity of the integrity assessment. In this study, piping failure probability when conducting and not conducting successive inspections using a probabilistic fracture mechanics (PFM) code, and the validity of successive inspections were examined quantitatively from the failure probabilities. The effect of crack sizing accuracy, which has great influence on the validity of inspections, was also examined. The validity of successive inspections was examined for 12, 16, and 24 inch pipes which were employed in PLR piping of BWR-5. It became clear that successive inspections were very effective in decreasing failure in 12 and 24 inch pipes, but not as effective in decreasing failure in 16 inch pipes. The reason is suggested that the crack growth rate in a 16 inch pipe is high compared with the other two sized pipes and opportunity of renewal of crack growth estimation at successive inspections is few. When successive inspections are carried out, there are several times of opportunity for improvement of underestimated crack growth behavior unless a crack size is measured smaller than the actual size in all successive inspections. Consequently, the influence of the crack sizing error on failure probability of pipes becomes very small.

This paper reports the results of the study on the failure modes and limit loads of piping in nuclear power plants subjected to cyclic seismic loading. By investigating the past fracture tests and earthquake resistance tests, it became clear that dominant failure mode of piping was fatigue, and the effect of ratchet strain was negligible. Until now, the stress generated with the acceleration of an earthquake was classified into the primary stress. However, the relationship between the input acceleration and the seismic response displacement of the pipe observed from earthquake resistance tests is non-linear, and increasing rate of displacement is lower than that of input acceleration in elastic-plastic stress condition. Therefore, the seismic loading can be treated as displacement controlled loading. To evaluate the reliability-based critical acceleration, a limit state function was defined taking the variations in the fatigue strength or some parameters into consideration. By using the limit state function, the reliability was evaluated for the typical piping of boiling water reactor (BWR) plants subjected to cyclic seismic loading, and a partial safety factors were calculated. Based on these results, a fatigue curve corresponding to the target reliability was proposed.

Negative pressure difference evaluation of Lungmen ABWR containment by using GOTHIC

The large capacity nuclear power plants have been widely constructed in recent years to cope with social economic development and electrical energy demands. Consequently, it is necessary to be aware of nuclear power plant (NPP) dynamic characteristics in order to study the transient and midterm dynamic process of power system. Due to lacking suitable NPP model which is described the detail characteristics of the Nuclear Power Unit in the power system simulation, at first, the paper presents functional description of nuclear power plant. Then a detailed pressurized water reactor (PWR) plant overall model for medium-term and long-term power system transient stability is proposed based on PSASP through user-defined modeling program. In the end, dynamic response characteristics of pressurized water reactor is presented. The simulation results are obtained.

The In-Vessel Retention (IVR) of molten core debris has been part of the severe accident management strategy for advanced pressurized water reactor (PWR) plant. Generally, the pressure is supposed to successfully be released, and the externally cooled lower head wall mainly experiences the temperature difference which may be more than 1000°C. However, the Fukushima accident shows that a certain pressure (up to 8.0MPa) still exists inside the reactor pressure vessel (RPV). Therefore, in order to make the IVR succeed, it is necessary to investigate the structural integrity of the RPV under the combined internal pressure and the thermal load on the lower head. Therefore, it is supposed that the lower head of RPV is a Hemisphere Shell Structure (HSS) with 2150mm external radius and 25mm wall thickness. Similarly, the outer wall temperature is supposed to be 127°C, the wall temperature difference 1200°C, and the material properties (i.e., thermal conductivity, specific heat capacity, yield strength, and so on) are dependent on the temperature. In this paper, the main subjects discussed are as below. First, the heat transfer analysis is carried out to obtain the temperature distribution of the HSS wall. Due to the high temperature gradient, there may present different failure modes along the HSS wall thickness, i.e., plastic failure and creep failure. Then, the stress and strain distributions along the wall thickness are analyzed by ANSYS finite element method where the Norton-Bailey type creep law is adopted. The result shows that the stress in the lower temperature part (LTP) whose temperature is lower than 405°C is lower than the yield stress of the material with the corresponding temperature. That is to say, the LTP still can bear a certain internal pressure. At last, to consider the combined impact of internal pressure and temperature difference on the HSS, the finite element analysis is carried out by adopting the isochronous stress-strain curves of the Chinese RPV material. The relationship between the stress in the LTP at 100 hours and the internal pressure is discussed, and the limit internal pressure is determined. It is concluded that HSS still can maintain its integrity under IVR, even if there exists a certain internal pressure.

Long-term constant load stress corrosion cracking (SCC) testing for alloys 690/152/52 at 360 ℃ is ongoing, showing no rupture for more than 105 h, suggesting immunity to primary water (PW) SCC initiation under stress level assumed for primary circuit components in pressurized water reactor (PWR) plants. Since the mechanical plug for steam generators has the largest cold work strain, a mock-up PWSCC test, using a mechanical plug of alloy 690, was also performed for evaluation of time to failure under stress and cold work conditions assumed for operating plan. As a result, it was proven that no crack initiated up to approximately 4 × 104 h at 360 ℃. PWSCC susceptibility was also evaluated in terms of crack growth rate (CGR). The CGR of alloy 690 increased after cold working, and the degree of increment is significantly affected by the nature of carbide precipitate along grain boundaries. It was found that increase in CGR caused by cold working remained relatively low when grain boundary carbides precipitated continuously along grain boundaries and coherently with the matrix. Contrarily, CGR grew higher in the materials with lower coherency. It was also revealed that alloy 690 with no grain boundary carbides (solution annealed alloy) showed a small increase of CGR after cold working.

10 The objective of this chapter is to discuss two approaches for reliability analysis of digital 11 instrumentation and control systems in nuclear power plants taking into account the 12 regulatory side. Dynamic Flowgraph Methodology (DFM) and Markov/Cell-to-Cell 13 Mapping Technique (CCMT) are discussed and case studies developed are presented. 14 These case studies involve simplified control systems for a steam generator and a 15 pressurizer of a Pressurized Water Reactor (PWR) plant for the purpose of evaluating 16 each method. Advantages and limitations of each approach are addressed. For the DFM 17 approach, three concerns in the literature are addressed: modeling of the system itself, 18 incorporation of the methodology results into existing Probabilistic Safety Assessments 19 (PSA), and identification of software failures. The Markov/CCMT, which has been used 20 in dynamic probabilistic safety assessments, is approached by means of a simplified 21 digitally controlled water volume control system. The Markov/CCMT methodology 22 results in detailed data of the system reliability behavior in relation to time. However, 23 it demands a higher computational effort than usual as the complexity (i.e., number of 24 components and failure states) of the system increases. As a regulatory research 25 conclusion, the methodologies presented can be used on PSA risk informed assessment, 26 contributing to the regulatory side. 27

piping, shrink, porosity, inclusions and microinclusions in the uranium alloy castings. In the completed fuel plates, a nonbond area of 1/8 of an ineh in diameter could be located. With the construction of a special two-arm probe, fuel plates installed in subassemblies were successfully tested by a transmission technique. (auth)

The nuclear renaissance that is emerging in the United States will be based on new power plant designs from a variety of US and international sources. While the new plant designs will be based on the time tested and proven concepts of both the boiling water reactor (BWR) and the pressurized water reactor (PWR), the emphasis of the new plants will be on engineering designs that incorporate increased levels of redundancy in safety systems, and post-LOCA passive systems to move fluids about the reactor containment, thereby providing for long-term core cooling and decay heat removal, The new plant designs will include a strong emphasis on modular, off-site construction where more protective coatings will be applied in the fabrication shop. This paper will investigate: (1) the anticipated use and types of protective coatings in new reactor designs, (2) relate how the NRC is moving to regulate the technical requirements applicable to coating selection and testing, 3) describe how the design philosophies inherent in the new reactor plants will potentially affect the performance envelope of the planned coating systems, and (4) discuss possible methodologies for the long-term maintenance of the applied coating systems.

Reactivity Initiated Accident (RIA) leads to an unwanted increase in fission rate and power in a region of the reactor core confined around the position of occurrence. The power excursion due to such events may cause fuel rods failures and a subsequent release of radioactive material into the primary coolant of reactor, in severe cases, this release could damage nearby fuel assemblies. In nuclear power plants, RIAs are due to control system faults, e. g. control elements ejection/insertion, or rapid changes in temperature or pressure of moderator. In Boiling Water Reactors (BWRs), the control rod drop accidents (RDAs) at cold zero power have been deeply investigated, in fact, notwithstanding they are less frequent in comparison with the control rod ejection event in PWRs, in this kind of plant these conditions are the most severe in case of a RIA occurrence. RDA transient, comprised in the design basis events considered in safety analysis, may cause rod failures especially at high burnup. To simulate a RIA, a peaked power pulse is applied to a pre-irradiated and re-instrumented rodlet aiming at investigating the most important phenomena that could lead to the rupture of cladding tubes. This paper is focused on the investigation of the TRANSURANUS fuel performance code capability to predict the thermomechanical state of rodlets subjected to RIA tests. To this purpose the FK-1 test, carried out at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute (JAERI), was simulated. This experiment is part of a set of 12 tests performed at the NSRR facility to study the performance under a reactivity initiated accident of BWR rodlets with burnup between 41 and 61 MWd/kgHM. In the FK-1 test, a STEP I BWR rodlet, previously irradiated in the Fukushima Daiichi Nuclear Power Station (Unit 3) operated by the Tokyo Electric Power COmpany (TEPCO) up to 45 MWd/kgHM, was subjected to a peak enthalpy insertion of 544 J/g. In this paper the code findings for the FK-1 test are discussed on the basis of the experimental data and the predictions of other stand-alone codes for transient analysis. The FK-1 predictions of FRAPTRAN (2001), FALCON (2003) and SCANAIR (ver. 3–2) are reported. The choice of fuel relocation model and important cladding properties (swelling, thermal expansion, thermal conductivity) was made relying on preliminary calculations whose results are also presented. Notwithstanding a satisfactory agreement between predictions and experimental data and a good agreement in the presented code-to-code comparison were envisaged, these results also emphasized the need to improve the models for FGR, heat transfer to plenum. Investigations are also required to ascertain possible contribution from fission gas to pelles thermal expansion. Ongoing modeling activity, performed at the ITU Joint Research Centre, is focused on a new model for FGR, ENEA (Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile) is expected, in the near term, to give a contribution to refine the model for plenum gas temperature. These activities should improve the description of RIA transient and further investigations on NSRR tests will be performed with newly developed models. The work presented in this paper will be part of ENEA contribution in FUMEX III project leaded by the International Atomic Energy Agency (IAEA) and aimed at the improvement of fuel codes predictions at high burnup.

In support of extended service and current operations of the US nuclear reactor plants, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating with Ginna Nuclear Power Plant, The Westinghouse Electric Company, LLC, and ATI Consulting, the selective procurement of baffle bolts that were withdrawn from service in 2011 and currently stored on site at Ginna. The goal of this program is to perform detailed microstructural and mechanical property characterization of baffle former bolts following in-service exposures. This report outlines the selection criteria of the bolts and the techniques to be used in this study. The bolts available are the original alloy 347 steel fasteners used in holding the baffle plates to the baffle former structures within the lower portion of the pressurized water reactor vessel. Of the eleven possible bolts made available for this work, none were identified to have specific damage. The bolts, however, did show varying levels of breakaway torque required in their removal. The bolts available for this study varied in peak fluence (highest dose within the head of the bolt) between 9.9 and 27.8x10²¹ n/cm² (E>1MeV). As no evidence for crack initiation was determined for the available bolts from preliminary visual examination, two bolts with the higher fluence values were selected for further post-irradiation examination. The two bolts showed different breakaway torque levels necessary in their removal. The information from these bolts will be integral to the LWRS program initiatives in evaluating end of life microstructure and properties. Furthermore, valuable data will be obtained that can be incorporated into model predictions of long-term irradiation behavior and compared to results obtained in high flux experimental reactor conditions. The two bolts selected for the ORNL study will be shipped to Westinghouse with bolts of interest to their collaborative efforts with the Electric Power Research Institute. Westinghouse will section the ORNL bolts into samples specified in this report and return them to ORNL. Samples will include bend bars for fracture toughness and crack propagation studies along with thin sections from which specimens for bend testing, subscale tensile and microstructural analysis can be obtained. Additional material from the high stress concentration region at the transition between the bolt head and shank will also be preserved to allow for further investigation of possible crack initiation sites.