Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). In RCP test facility (RCPTF), it is possible to perform a type test for a newly-developed RCP as well as a production test for the Advanced Power Reactor 1400 (APR1400) RCP before its installation in nuclear power plants. The thermal hydraulic and electric capability of the RCPTF covers up to 18.5 MPa, 343 °C, 11.7 m 3 /s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. During the commission test period, the technical activity for RCP performance verification test and optimization of flow stability have also been performed. The developed techniques can be divided into four subjects; 1) Development of flow stabilization technique for high flow condition test facility, 2) Improvement on the accuracy of performance verification factor and development of evaluation method, 3) Evaluation of test facility vibration characteristics and optimization of measurement system, and 4) RCP coast-down data production. KAERI has completed a full set of technique developments, a prerequisite for the RCP performance test.

The US NRC’s post-Fukushima taskforce recommend extending the coping time up to at least 72 hours. In PWRs, the reactor coolant pump (RCP), and more particularly the shaft seal system located between the motor and the impeller, are directly impacted by these recommendations [AREVA, 2012]. Considering the importance of station black out (SBO) condition, seal performance of RCP shaft seal assembly shall be verified especially under an extended SBO condition. As a final stage of the localization program of the RCP, KAERI performed a seal LOCA test to verify seal leakage characteristics and to produce qualified experimental data for a plant safety analysis during the extended SBO condition. The seal verification test was performed using the localized seal assembly, fabricated by DOOSAN Heavy Industries & Construction Company of Korea, along with the commercial APR1400’s RCP for the Shin-Kori unit-5 nuclear plant to simulate an actual condition of the commercial nuclear power plant. Test condition was derived from a series of safety analysis for the APR1400 plant which reflected the FLEX procedure, and it consisted of three major sub-conditions such as high-pressure and high-temperature condition, saturated condition, and low-pressure and low-temperature condition. The major concerning point of the test is to measure a seal leakage flow rate during the test period. Total leakage amount and maximum leakage flow rate through the localized shaft seal assembly during the whole period of the test are in comparatively smaller ranges than the conventional input values used for a safety analysis of APR1400.

Dynamic response research of steam generator loss of coolant accident (SG LOCA) is essential for the reliability and safety consideration. According to the differences of LOCA loading phenomena, two types of LOCA loads affect the SG: rarefaction wave travels through the primary fluid in the U-tubes, and the SG shakes due to reactor coolant loop(RCL) motions transmitted by the primary loop piping, former loading phenomena evaluation is called SG rarefaction analysis while latter is called shaking analysis. This paper place particular emphasis on shaking analysis. At present, the published literatures about LOCA mainly focus on RCL LOCA, reactor LOCA and fuel assembly LOCA, few reports concentrate on shaking dynamic response analysis of SG LOCA. Both Westinghouse and AREVA’s methods according to their research reports are to decouple the SG from the RCL: This method results in low computational efficiency as RCL LOCA and SG LOCA are evaluated separately and the decoupling error is uncertainty, meanwhile, in the vicinity of the nodes where the displacements are imposed, distorted reaction forces are usually found. Through reasonable simplification and equivalence, a detailed nonlinear FEM model of steam generator (SG) of a China 3rd generation nuclear power plant (NPP) is established, this model is then connected with the reactor coolant loop (RCL) to carry out the SG LOCA shaking dynamic response analysis. By calculation, the maximum absolute stresses of SG heat transfer tube bundle and its variation with tube diameter and upper supports reacting forces are obtained. In order to study the effect of SG decoupling from the RCL on shaking dynamic response, a comparative study of decoupling /coupling methods is conducted. Results shows that SG decoupling has a significant impact on the calculation result, the calculation method of coupling is more closer to the real situation and worthy to recommended. Related analytical procedures and calculation results lay the foundation for future SG shaking dynamic analysis and SG design of subsequent power plants.

High confidence of low probability of failure (HCLPF) values of equipment, representing the seismic capacities of the equipment, are the fundamental ingredient in seismic probability safety assessment (SPSA) and seismic margin analyses (SMA). In this paper, two methods for calculating the HCLPF values of equipment were investigated, fragility analysis, and conservative deterministic failure margin (CDFM). These methods are linear methods. Based on these methods, HCLPF value of equipment can be computed conveniently by scaling the results of the existing seismic analysis. For a nonlinear systems, the HCLPF values based on these linear scaling methods are unrealistic. For a complicated nonlinear equipment or structure, a detail nonlinear model was used to derive the seismic capacity. The results by this method are realistic, but cost calculation time. In this paper, a nonlinear model of reactor coolant system coupled reactor building was built. This model includes the steam generator and considers the nonlinear factors of steam generator such as gap in the supports, plasticity of hot leg and cold leg. Forced motion was applied to the base of reactor building. And seismic response of the steam generator was calculated iteratively by scaling the ground motion level step by step. Based on these calculations, a curve of load on the supports VS peak ground acceleration (PGA) can be obtained. Then based on these curves and allowable load of supports of steam generator, which derived from stress analysis on support of steam generator, seismic capacity of the supports of steam generator was determined. Then the HCLPF Value of the supports of steam generator was obtained by this nonlinear time history analysis and was compared with the results based on the CDFM. The two results were different. Therefore, the HCLPF seismic capacity of equipment with nonlinearity, such as gap nonlinearity, should be calculated by nonlinear time history method.

Steam generator tube rupture (SGTR) accident is one of important accident that has high probability of resulting in severe accidents. As a bypass scenario, fission product can be directly released to the environment during the SGTR accident. Thus, the severe accident by SGTR should be carefully managed by severe accident management guidance (SAMG). In Korea, SAMG for optimized power reactor 1000 (OPR1000) has been developed in 1999 and used to mitigate the severe accident of OPR1000 with seven mitigation strategies. Among the mitigation strategies, ‘Depressurization of reactor coolant system (RCS)’ is one of the most powerful strategies to reduce direct release of the fission product. To reduce the RCS pressure, indirect depressurization using steam generator is generally recommended. However, depending on the RCS condition, the indirect depressurization can be ineffective to reduce the RCS pressure. In this case, direct depressurization using pilot operated relief valve (PORV) should be performed as a second plan. From this point of view, sensitivity study of RCS depressurization was performed to investigate priority of depressurization in this study. The severe accident scenario initiated by SGTR accident was selected from probabilistic safety assessment (PSA) level 1 report and simulated using MELCOR 2.1. For the mitigation strategy, various timing of depressurization, the number of opening valves and flow rate of feed water were applied to simulate the possible depressurization strategies during the severe accident. The MELCOR code simulation shows that if depressurization was performed at 30 minutes after SAMG entrance, the direct depressurization was more efficient to reduce the RCS pressure and the fission product release. Therefore, it was recommended to use direct depressurization rather than indirect depressurization in certain time. The sensitivity of flow rate of feed water and different number of opening valves were insignificant for progress of the accident and fission product release. In conclusion, operators should select the way of depressurization to reduce the RCS pressure and the fission product release during the SGTR accident, considering the condition of the plant such as accident progress and availability of safety features. To suggest more proper information for depressurization, more sensitivity analysis and detailed thermal-hydraulic analysis should be performed for the future work.

Multi-objective optimization is a powerful tool that has been successfully applied to many fields but has seen minimal use in the design and development of nuclear power plant systems. When applied to design, multi-objective optimization involves the manipulation of key design parameters in order to develop optimal designs. These design parameters include continuous and/or discrete variables and represent the physical design specifications. They are modified across a specific design space to accomplish a number of set objective functions, representing the goals for both system design and performance, which conflict and cannot be combined into a single objective function. In this paper, a non-dominated sorting genetic algorithm (NSGA) and parallel processing in Python 3 were used to optimize the design of the passive endothermic reaction cooling system (PERCS) model developed in RELAP5/MOD 3.3. This system has been proposed as a retrofit to currently-operating light water reactors (LWR) and is designed to remove decay heat from the reactor core via the endothermic decomposition of magnesium carbonate (MgCO 3 ) and natural circulation of the reactor coolant. The PERCS design is currently a shell-and-tube heat exchanger, with the coolant flowing through the tube side and MgCO 3 on the shell side. During a station blackout (SBO), the PERCS initially keeps the reactor core outlet temperature from exceeding 635 K and then reduces it to below 620 K for 30 days. The optimization of the PERCS was performed with three different objectives: (1) minimization of equipment costs, (2) minimization of deviation of the core outlet temperature during a SBO from its normal operation steady-state value, and (3) minimization of fractional consumption of MgCO 3 , a metric that is measurable and directly related to the operating time of the PERCS. The manipulated parameters of the optimization include the radius of the PERCS shell, the pitch, hydraulic diameter, thickness and length of the PERCS tubes, and the elevation of the PERCS with respect to the reactor core. The NSGA methodology works by creating a population of PERCS options with varying design parameters. Using the evolutionary concepts of selection, reproduction, mutation, and survival of the fittest, the NSGA method repeatedly generates new PERCS options and gets rid of less fit ones. In the end, the result was a Pareto front of PERCS designs, each thermodynamically viable and optimal with respect to the three objectives. The Pareto front of options as a whole represents the optimized trade-off between the objectives.

During the reflood phase of a postulated large break loss-of-coolant accident (LBLOCA), the liquid head in the reactor vessel downcomer provides the driving force to reflood the core. Since the reflood rate is a function of the downcomer inventory, the calculation of the downcomer liquid inventory is critical in simulating the reflood phase of a postulated LBLOCA accident in a pressurized water reactor. Since the reactor coolant system pressure decreases rapidly after the onset of a LBLOCA transient, the walls surrounding the downcomer become superheated for the duration of the transient. The Japan Atomic Energy Research Institute (JAERI) downcomer effective water head test facility was designed to study boiling and steam-water interaction in the reactor vessel downcomer under prototypical reflood conditions. A number of tests were conducted at this facility with varying degrees of wall superheating (among other things) that cover the expected degree of superheating in a pressurized water reactor. The wall superheating achieved at the JAERI facility is greater than that of other large-scale facilities that are typically simulated to validate thermal-hydraulic system codes. W COBRA/TRAC-TF2 is the thermal-hydraulic system code utilized in the FULL SPECTRUM ™ LOCA ( FSLOCA ™) evaluation model (EM). The ability of the W COBRA/TRAC-TF2 code to predict phenomena occurring in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA has been previously validated. However, only limited wall superheating was present in the existing validation basis. As such, two experiments conducted at the JAERI downcomer effective water head test facility are simulated to provide additional information on the capability of W COBRA/TRAC-TF2 to predict the liquid inventory in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA. The code captured all the trends observed in the experimental data for both Run 115 and Run 121. The various collapsed liquid levels tended to be well-predicted or under-predicted by the code after the initial simulated accumulator injection period.

For PWR, remote shutdown station (RSS) is a redundant control mean to shut down the reactor when main control room (MCR) inhabitation is challenged (e.g. fire, smoke...). Nowadays, due to nuclear power plants control measures were improved with DCS system, a full function DCS RSS was equipped and more essential equipment could be controlled on RSS. Under operating conditions that prohibit nuclear power plant operators to stay in the main control room, the operators should move to RSS and shutdown the reactor to ensure plant safety following <Moving to remote shutdown station when main control room is un-inhabitable operating strategy> (RSS strategy for short) to fallback the plant from power operation to cold shutdown. The original operating strategy by nature circulation is no longer the best choice both for operation safety and economy efficiency, and an optimized new strategy should be raised. Based on the former reason, an optimized operation strategy was raised in this paper. In the optimized strategy, all plant normal standard operation modes were considered as initial conditions, rather than only considering power operation condition in the original one. The fallback mode and fallback strategy for each initial condition was also designed and optimized. To accelerate the depression and heat removal process, a forced circulation operation strategy is adopted when the reactor coolant pumps are available, and less local operation was included by taking advantages of the full function operating measures on RSS. To simplify the whole procedure structure, the operation modules of other general operating procedures are reused. To validate the effectiveness of the optimized operating strategy, a full scope PWR simulation tool was employed to make thermo hydraulic calculation validation of the reactor response and also the remote control station HMI supporting validation. By simulating the original strategy and the optimized one and related analysis, we found that the optimized strategy is effective, and able to be executed based on the remote control station hardware. By executing the optimized strategy, the unit can fall back to the cold shutdown condition safely and a few hours were saved compared with the original strategy. The optimized strategy had already been implemented on real PWR nuclear power plant.

Pressurized Water Reactor (PWR) nuclear power plant sump operator assisted program is applied to monitor unrecognized leaks of reactor coolant. It is very crucial to leak before break (LBB) protection and greatly affects the operational safety of nuclear reactors. In this paper, an algorithm of sump level operator assisted support program is proposed. Compared with the algorithm of traditional PWR, this algorithm adds the identification of working conditions and re-builds the leakage flow calculation method, which eliminates interference factors to the extent practical and improves the accuracy of the calculation results of unrecognized leakage flow.

“Begging the question” describes a situation in which the statement under examination is assumed to be true (i.e., the statement is used to support itself). Examples of this can be found in analysis reports that were prepared by analysts who are not mindful (or maybe uninformed) of the analysis criteria they’re required to fulfill. This is generally seen in analyses of anticipated operational occurrences (AOOs). AOOs are defined in Appendix A of 10 CFR §50 [1], and in ANS-N18.2-1973 [2], where they’re also known as American Nuclear Society (ANS) Condition II events. This standard [2] also defines more serious, Condition III and IV events. Analyses of AOOs, or ANS Condition II events are required to show that: (1) reactor coolant system (RCS) pressure will not exceed its safety limit, and (2) no fuel damage will be incurred, and (3) a more serious accident will not develop, unless there is a simultaneous occurrence of another, independent fault. The three requirements are often demonstrated by three different analyses, each of which is designed to yield conservative results with respect to one of the requirements. Accident analyses that are performed to demonstrate compliance with the first two requirements are relatively straightforward. They rely mostly upon the design of safety valves and the timing of reactor trips. “Begging the question” is seen in analyses that are designed to demonstrate compliance with the third requirement. This paper will describe how this logical fallacy has been applied in licensees’ accident analyses, and accepted by the NRC staff.

Water chemistry plays an important part in maintaining corrosion resistance in water transport systems throughout nuclear power plants (NPP’s). Small changes in liquid chemistry such as pH, borate concentration, or build-up of crud in reactor cooling water can result in rapid degradation or damage to components and lead to unexpected failures. The Chemical and Volume Control System (CVCS) and Reactor Water Cleanup System (RWCU) are responsible for maintaining these parameters at appropriate levels, and so failure of either of these systems can result in unnecessary stresses on many other reactor systems due to resulting transients. While the major components of these systems all have sufficient redundancy to prevent major accidents, failure of components in these systems can result in failure of other redundant components and affect plant safety [1]. The CVCS and RWCU systems have experienced aging related degradations and failures in the past, and although they have not affected the system’s emergency functions, they have resulted in unnecessary actuation of related systems, and reactor shutdowns [1]. Reactor shutdowns can result in large changes in reactor coolant chemistry such as oxygen and borate concentration transients, and the build-up of corrosion products which can’t be as easily removed during periods of reactor shutdown [2]. In the following analysis of Component Operational Experience Degradation and Ageing Program (CODAP) experience data; causes, impacts, and preventative actions as recorded in CODAP are examined for degradation events which took place in the CVCS and RWCU, of PWRs and BWRs, respectively. The analysis will demonstrate the usefulness of CODAP in examining reactor component failure trends, as well as discuss insights on improvement for the program.

The vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs. Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle. Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N 2 mixture, application of SuperCritical (SC)-CO 2 cycles or combined cycles. The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO 2 is provided.

The natural circulation small modular reactor (NCSMR) is a 330 MW reactor which has no reactor coolant pumps (RCP) and no active safety injection systems at all. The reactor is mainly comprised of the reactor pressure vessel (RPV) with integral pressurize r and steam generator. RPV is enclosed by a vacuumed pressure containment vessel (PCV) and the PCV is submerged in the underground containment pool. A MELCOR model and corresponding input deck are developed for the RPV, PCV, and containment pool. The containment pool takes the role of ultimate heat sink (UHS) in accident situations. The containment pool may crack and leak in some critical accidents as the earthquake, leading to the severe accident of the reactor. A TMI-2 like SBLOCA in the RPV (stuck open RVVs) along with the containment pool crack (loss of ultimate heat sink) is simulated in the work. So me key parameters as the RRVs stuck open fraction, the PCV-SRVs open or not, the containment pool crack position would have large influence on the severe accident sequence. The sensitivity of these parameters to the accident sequence is analyzed in the work. According to the simulation results, the RPV pressure decreased with the RRVs stuck open. The depressurization of RPV accelerated with the RPV-SRV open fraction increase. The PCV pressure increased after that. Two cases as the PCV-SRV open after PCV pressure increase to 5 MPa, and PCV break while the RV d id not open, are analysis. The coolant discharge mass flo wrate in RPV and PCV were different in two cases, leading to the different degradation situation of the core. Since the containment pool is so important for the accident mitigation, sensitivity analysis is done for the containment pool crack position in the pool. The work will be meaningful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents during the design process of NCSMR.

Focused on the overall experimental verification of important phenomena in small break loss of coolant accident (SBLOCA) of large-scale passive PWR plants, this paper put forward a top-down and bottom-up scaling analysis of SBLOCA process based on H2TS method, by which the scaling conditions of IET facility would be figured out to match with the conditions in real nuclear power plants. Meanwhile, parameter evaluation was conducted between IET facility and real power plants, and finally the results proved that the IET facility could, to a certain degree, represent the conditions in real plants in terms of scaling analysis, and the test data of facility could be applicable for the SBLOCA of real plants. Further, in this paper, scenarios of SBLOCA were simulated and compared between IET facility and AP1000 nuclear power plant by the use of Relap5/MOD3 program. Then, after the conversion of flow area and time ratio, the comparison of results between those two systems was conducted in terms of pressure of reactor coolant system (RCS), injection flow rate of CMT and steam flow rate through ADS-1/2/3/4 valves, which also proved that the IET facility could simulate the SBLOCA of real plants well.

The loss of secondary heat sink accident caused by loss of all feed water for a PWR with medium pressure safety injection system (MPSI) design is analyzed in order to support the development of one of the emergency operating procedures (EOPs), “response to loss of secondary heat sink”. The bleed and feed (B-F) measure to mitigate the accident is analyzed in detail. For the bleed and feed measure, the safety injection flow is established only after the reactor coolant system (RCS) pressure is lower than the pump head of MPSI, which needs the open of the pressurizer (PRZ) safety valves to depressurize the RCS ahead. To prevent the core uncovery or minimize the fuel cladding temperature during the bleed and feed process, the time and the number of PRZ safety valves needed to be opened are investigated. A best estimated plant model with RELAP5 code is applied for the sequences calculation. The result shows that the core water level and fuel cladding temperature will be benefited from an earlier open of all safety valves.

Reactor coolant pump is one of the key equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor reactor coolant pump, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a clearance flow. The fluid induced forces of the clearance flow in canned motor reactor coolant pump and their effects on the rotordynamic characteristics of the pump are experimentally analyzed in this work. A vertical experiment rig has been established for the purpose of measuring the fluid induced forces of the clearance. Fluid induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor reactor coolant pump does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid induced forces of the clearance flow.

According with the requirements of HAF-102[1], the purpose of radiation protection optimization is to achieve occupational exposures as low as reasonably achievable (ALARA). More than 80% of the collective exposure doses are received during the outage for PWRs. Activated corrosion products, especially Co-60, deposited on the surface of reactor coolant system are the main causes of these doses. Co-60 is the principle contributor to out-of-core radiation fields in PWRs, which is a high energy gamma emitter with a 5.3 year half-life period. The contamination by Co-60 plates out in plant piping systems and produces long term high radiation areas in the plant. Stellite, a hard facing alloy trade, which has high Co-59 content (approximately 60%), is identified as one of the most important source of Co-60 in reactors, whereas the contribution of Stellite to cobalt are always not paid enough attention to. Cobalt is released through ware and corrosion of Stellite. When cobalt is released in systems with a flow path to the reactor, it can become activated by neutron to Co-60. This paper analyzes the Stellite material release into the primary coolant. Based on the above considerations, the impact of Stellite replacement on plant contamination by Co-60 is calculated and analyzed with the use of activated corrosion product source term calculation program. The primary conclusions are as follows: (1) Reducing the quantity of Stellite material used during the design and building of PWRs can significantly reduce the Co-60 deposition on primary circuit. (2) The replacement of Stellite material has a positive impact on PWR contamination by Co-60, but the outcome is not so significant. Based on the above analysis, recommendations are made. Stellite material replacement project can be used for the operating reactors to reduce the Co-60 contamination. But for the design and building of new reactors, reduction of the use of Stellite material is the best choice.

The main equipments of the reactor coolant system include the steam generator, the reactor coolant pump, the pressurizer and the reactor coolant loop. The reactor coolant system is equipped with a steam generator for each of the two loops, and pressurizer is connected with the hot leg of loop 1 using the surge line. The main loop support system design of AP series greatly simplifies the RCS loop support system. Pressurizer supports consist of columns, lower lateral supports, upper lateral support and ring girder, and the steam generator supports consist of columns, lower lateral supports, upper lateral supports and intermediate lateral supports. Ring Girder of pressurizer consists of two semi-circular girders, vertical supports and splice connection of girder and the two half-ring girders are connected with splice connection using 11 bolts. The steam generator upper lateral support is mainly composed of bracket, snubber, pin and ubar and the ubar and the steam generator is connected via 16 bolts in the initial design. These bolts are to ensure the support junction can withstand the force and torque of various conditions of the reactor coolant system, which are important components of the main equipment support. There are large numbers of bolts in the splice connection of ring girder and ubar of upper lateral support of steam generator, and the bolts load was calculated using the uniform method in the general engineering design and analysis. During the design review it was found that the bolts load was uneven and in order to determine the non-uniformity of the bolts the finite element method was used to calculate the load on each bolt, and the resulting stress ratio was greater than 1 did not meet the requirements of the ASME Code. In this paper, the calculation method was studied and the design improvements for parts was made using the nonlinear analysis method to meet the requirements of ASME Code in the case of master supports of main equipment supports have been made good. At the same time the impact of bolts load because of gap was studied. It had very good economic benefits. The calculation and research of this paper show that the finite element method can calculate the force of bolts finely, and can get a more reasonable result than the empirical formula. It can be referred to when the multi-bolt connection structure needs to be refined, such as flange connection of important equipments and valves and flange and bolt design optimization.

Accumulative test data indicates that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of materials used in the reactor coolant pressure boundary components to significantly reduce. EAF is used as the abbreviation of the environmentally assisted fatigue in the nuclear field. In 2007, NRC issued RG. 1.207. It was updated in 2014. And it requires that the effects of the light-water environment on the fatigue life reduction of metal components should be considered for new plants. And it suggests to use environmental correction factor (Fen) to account for EAF. Fen = Nair/Nwater (N is occurrences). NUREG/CR-6909 [1] presents the detail Fen calculation formula which includes the complicated influence of combined multi-parameters. Fen is a function of temperature, strain amplitude & rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depends on the experience of the primary pressure boundary piping transient operation, Fen vary during each transient. More uncertainty and confusion are raised during the application of the Fen method. In the research work involved in this article, first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depend on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multi-transient loading conditions. And the result is compared with that of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test matrix is needed to prove its practicability furthermore.

Generic Safety Issue 23 (GSI-23) by the U.S. Nuclear Regulatory Commission (NRC) was identified in 1980 as a result of staff concerns about reactor coolant pump (RCP) seal failure during Station Blackout (SBO), that is, seal degradation leading to a significant loss of reactor coolant in pressurized-water reactor (PWR) plants. Resolutions of GSI-23 have been considered at PWR plants. In 2000, NRC decided to close GSI-23 and issue Regulatory Issue Summary00–002 (RIS 00–002) [1] , based on considerations such as; for example, the improvement of RCP seal performance [2] , and the reduction of the risk of RCP seal failure in certain plants by the addition of alternate power sources. After the closure of GSI-23, some licensees were planning to make other associated improvements under their individual plant program. In Japan, the RCP seal was showed that leakage rate was low under SBO testing conditions [3] in licensing safety reviews conducted according to new nuclear regulatory standards after the nuclear accident at the Fukushima Daiichi Nuclear Power Plant. On the other hand, a boiling-water reactor (BWR) is not included in GSI-23 because operating experience indicates that seal failures in BWRs result in smaller leak rates than seal failures in PWRs and, BWRs have the reactor coolant injection capability under SBO conditions, such as the reactor core isolation cooling (RCIC) and the high-pressure coolant injection (HPCI) system. In addition, for the particular BWR-2 type plants that do not have emergency makeup systems, the pump mechanical seal was tested under SBO conditions and successfully showed minimal leakage [4] . However, for the BWR-5 type plants which have the reactor coolant injection capability, such as the RCIC and HPCI systems, the pump seal had not been tested. In Japan, after the nuclear accident at the Fukushima Daiichi Nuclear Power Plant, licensing safety reviews of BWRs and PWRs are being conducted according to new nuclear regulatory standards. We took this opportunity to test the leak rate from Primary Loop Recirculation (PLR) pump mechanical seals under SBO condition. The peak of leak rate was approximately 0.6ton/h (2.6gpm) during the 24 hours of SBO testing condition. Despite damage of O-rings in the mechanical seal by heated water which were ovserbed at post-inspection test, a very low leak rate was realized because the leakage path after passing through the damaged parts of the O-rings was limited by the other restricting pathway in the mechanical seal. This seal leakage was very low, compared with the reactor coolant makeup capability of the RCIC system and the reactor coolant release capability from main steam safety relief valve (SRV). Therefore, we reconfirmed that the result of this leak rate does not affect the safety evaluation for the reactor. It is shown in this paper that the leak rate from PLR pump mechanical seals is low under SBO condition by our demonstration test.