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Although numerical welding simulation is now commonly used in the nuclear industry to predict residual stresses in reactor vessels and associated piping components, there are currently no universally accepted guidelines for performing such analysis. Moreover, due to the complexity of the calculations and varying analytical procedures among analysts, there remains a need to validate predictions of residual stress against benchmark studies. As part of an industry initiative to manage the degradation of dissimilar metal welds in pressurized water reactor piping that are susceptible to primary water stress corrosion cracking, the U.S Nuclear Regulatory Commission embarked on a multi-phased program to validate welding residual stress models. The aim of Phase II of this program is to obtain measured residual stresses from a pressurizer surge nozzle dissimilar metal weld mockup for use in comparisons with numerically predicted stresses. This paper presents results of finite element analysis for various stages during the fabrication of a 14–inch pressurizer surge nozzle mockup, including an Alloy 82 dissimilar metal weld between a stainless steel safe end and carbon steel nozzle, an inside surface weld repair (back weld) and fill-in weld (weld build-up), and a stainless steel “field” weld attaching a section of straight pipe to the safe end. The NRC validation program was structured to allow participants to first calculate results using their own material properties, and then tune their welding simulations to thermocouple data. This was followed by reanalysis using NRC-supplied material properties. The program was conducted as a round robin analysis among an international group of participants and formatted as a blind validation project wherein results were submitted to the NRC prior to receipt of thermocouple and material property data. Results were obtained for both kinematic and isotropic hardening rules to study the effect of these two extreme measures of material characterization on the development of residual stress. Predicted stresses are then compared to measured stress data obtained by the deep-hole drilling technique at multiple locations through the thickness of the weld. The NRC residual stress model validation project serves as a valuable contribution to the understanding of how residual stresses are developed in dissimilar metal welds. The correlation of calculated residual stresses with measured data from a relevant mockup also serves to increase confidence in predicting crack growth in these primary pressure boundary welds by removing much of the uncertainty previously associated with residual stress input to crack growth analysis.Copyright © 2011 by ASME

In November 2009, the Electric Power Research Institute (EPRI) assembled an Advisory Panel on Environmental Fatigue, consisting of various industry and vendor participants, whose charter is to provide insight and recommendations to EPRI (including the Pressurized Water Reactor (PWR) Materials Reliability Program (MRP), the Boiling Water Reactor Vessel and Internals Project (BWRVIP), and the Advanced Nuclear Technology (ANT) organizations) on conducting additional work and providing guidance associated with environmental fatigue. The goal of the panel is to identify areas in environmentally assisted fatigue (EAF) formulations, or application of the EAF formulations, contained in regulations and the ASME Code that may cause application difficulties, to determine where new research efforts are necessary to provide guidance or alternatives to application of these formulations, and to investigate industry design problems with a goal of identifying solutions to obtain near- and long-term relief. The Advisory Panel consists of various industry and vendor participants that have been engaged in the environmental fatigue topic over the past several years. As a part of the activities investigated by the Expert Panel, in 2010, a sample problem was defined for use in testing the application of proposed ASME Code Cases for evaluation of EAF using the EAF multiplier (Fen ) correction factor methodology. The sample problem was provided to multiple industry organizations and regulators to be solved so that differences in approaches and methods applied to the sample problem could be identified and understood. The intent of this effort was to refine the Fen methodology to clarify issues that have been previously identified and its practical application to typical industry fatigue evaluation problems. NRC staff elected to participate in solving the sample problem and comparing solutions generated by the other participating organizations in order to more fully understand the issues with fatigue calculation methodologies, as well as to provide input and develop consensus on the evaluation approaches. The sample problem represents a simplified form of a traditional reactor pressure vessel-type evaluation for transient thermal stress analysis and fatigue cumulative usage factor (CUF) calculation of a representative piping component evaluated in accordance with ASME Code, Section III, Subarticle NB-3200 methods. This paper describes detailed investigation of differences between two solutions to the Expert Panel sample problem. This investigation involves comparisons of results from the initial finite element solution for thermal and pressure stresses (using different finite element software packages), to the calculation of CUF (both with and without EAF effects). Observations regarding differences in results and the reasons for those differences are identified and discussed. Recommendations for eliminating these differences are also made.Copyright © 2011 by ASME

Several Ni-base alloys and their weld metals such as Alloy 600 and Alloy 82/182 suffered from stress corrosion cracking in pressurized water reactor primary water environments. Materials Reliability Program (MRP) proposed a CGR disposition curve in a report MRP 55 for PWSCC of thick-section Alloy 600 materials. This deterministic CGR equation has been adopted by Section XI Nonmandatory Appendix O of the ASME Boiler and Pressure Code for flaw evaluation. MRP also proposed a CGR disposition curve in MRP report 115 for PWSCC of Alloy 82/182/132 weld metals. In the same fashion, JSME and JNES also provided CGR disposition curves in the flaw evaluation procedure in structural integrity analysis. Stress intensity factor (K), temperature and thermal activation energy are included in both MRP 55 and MRP 115 reports. Both MRP 55 and MRP 115 are engineering-based rather than mechanism-based. The fundamental correlations such as crack growth rate vs. K are quantified based on the theoretical model and screened experimental data, which are compared to the reported disposition curves and used for improving the prediction.

The Korea Atomic Energy Research Institute (KAERI) has been developing a dual-cooled annular fuel for the power uprate of 20% in an optimized pressurized water reactor (PWR) in Korea, OPR1000. The dual-cooled annular fuel is configured to allow the coolant flow through the inner channel as well as the outer channel. Several thermal-hydraulic issues exist for the application of the dual-cooled annular fuel to OPR1000. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually results in a fuel failure. A long lower end cap for the annular fuel was invented to provide flow holes by perforating the side surface of the end cap body. The side holes in the lower end cap are expected to supply a minimum coolant in the inner channel in order to prevent the DNB occurrence in the event of partial or even complete blockage of the inner channel entrance. But due to very unusual shape of the lower end cap, it is difficult to estimate the flow resistance of the side flow holes using empirical equations available in the open literatures. Experimental and computational fluid dynamics (CFD) study were performed to investigate the bypass flow through the side holes of the end cap in the case of complete entrance blockage of the inner channel. The form loss coefficient in the side holes was also estimated by using the pressure drop along the bypass flow path.Copyright © 2011 by KSME

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress (WRS) predictions in pressurized water reactor (PWR) primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in PWRs are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are the primary driver of this degradation mechanism. The NRC/EPRI weld residual stress (WRS) analysis validation program consists of four phases, with each phase increasing in complexity from laboratory size specimens to component mock-ups and cancelled-plant material. This paper discusses Phase 2 of the WRS characterization program involving an international round robin analysis project in which participants analyzed a prototypic reactor coolant pressure boundary component. Mock-up fabrication, WRS measurements and comparison with predicted stresses through the DM weld region are described. The results of this study show that, on average, analysts can develop WRS predictions that are a reasonable estimate for actual configurations as quantified by measurements. However, the scatter in predicted results from analyst to analyst can be quite large. For example, in this study, the scatter in WRSs through the centerline of the main DM weld (prior to stainless steel weld application) predicted by analysts is approximately +/− 200 to 300 MPa at 3 standard deviations for axial stresses and +/− 300 to 400 MPa at 3 standard deviations for hoop stresses. Sensitivity studies that vary important parameters, such as material hardening behavior, can be used to bound such large variations.Copyright © 2011 by ASME

Welding residual stress modeling is currently performed by researchers around the world using a wide variety of modeling methods to predict the final stress state of a completed weld. Among the key modeling assumptions used to perform a residual stress simulation are: • The geometric setup of the model, including boundary condition and assumptions. • The thermal and structural model simulating the welding process including lumping of weld passes. • The strain hardening input properties and hardening law. Researchers from Dominion Engineering, Inc. (DEI) and Westinghouse Electric Co. (WEC) have performed a benchmark comparison studying these key modeling assumptions and their results on the predicted welding residual stress distributions. Researchers from DEI and WEC have completed independent studies to validate their respective methods for calculating residual weld stress. In addition to the comparative evaluation, brief descriptions of the individual validations will be included in this paper. The weldment selected for evaluation is a typical reactor pressure vessel (RPV) outlet nozzle dissimilar metal safe end weld in a pressurized water reactor plant. This weld joins a low alloy steel nozzle to a stainless steel safe end using Alloy 182 weld material; this weld is completed in the manufacturing shop. The safe end is then field welded to the stainless steel reactor coolant loop piping. The residual stress distributions in the dissimilar metal welds, like the one selected, are important in predicting stress corrosion crack growth in Reactor Coolant System (RCS) components. The fabrication drawings for the selected RPV outlet nozzle were provided to both organizations, and independent residual stress simulations were performed using the best effort modeling techniques from each organization. This paper investigates the impact of the key modeling assumptions described above on the differences in the predicted welding residual stress distributions between the two simulation techniques. The results from the modeling comparison are provided in this paper.Copyright © 2011 by ASME

The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI welding residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material. This paper discusses Phase III of the WRS characterization program, comparing measured and predicted weld residual stresses profiles through the dissimilar metal weld region of pressurizer safety and relief nozzles removed from a cancelled plant in the United States. The DM weld had already been completed on all of the plant nozzles before use in the mock-up program. One of the nozzles was completed with the application of the stainless steel safe-end weld to a section of stainless steel pipe. Measurements were taken on the nozzles with and without the welded pipe section. Several independent finite element analysis predictions were made of the stress state in the DM weld. This paper compares the predicted stresses to those found by through-thickness measurement techniques (Deep Hole Drilling and Contour Method). Comparisons of analysis results with experimental data will allow the NRC staff to develop unbiased measures of uncertainties in weld residual stress predictions with the goal of developing assurances that the analysis predictions are defensible through the blind validation provided using well controlled mock-ups and ex-plant material in this program.Copyright © 2011 by ASME

HI-SMUR 140 is a small (145 MWe) modular pressurized water reactor designed to harness fission energy without the use of a recirculation pump. HI-SMUR’s core resides deep underground in a thick-walled reactor vessel (RV) enclosed by a stainless steel-lined reinforced concrete “Reactor Well”. The HI-SMUR Nuclear Steam Supply System (NSSS) is a conjugated pressure vessel assemblage wherein the steam generator(s) are integrally joined to the RV, i.e., without any interconnecting piping, and the pressurizer is integral to the reactor vessel. There are no penetrations in the RV for over a height of 120 feet above the reactor core, which precludes the scenario of loss of coolant to the core from a postulated pipe break event. The rejection of decay heat from the reactor in the wake of a scram is engineered to occur without the aid of on-site or off-site power, making the HI-SMUR NSSS demonstrably capable of withstanding a cataclysmic environmental phenomenon of Fukushima’s intensity without loss of core cooling or without precipitating any damage to the plant or the surrounding environment. The system design places a premium on accessibility and maintainability of vital equipment such as the steam generators, RV internals, and the control rod drive assemblies. This paper is the first in a series of papers planned to explain and quantify the performance and safety aspects of HI-SMUR 140. In this paper, the thermal-hydraulic characteristics of the HI-SMUR NSSS are explored using classical hydraulic correlations which have served as the tool for the scoping parametric study of the system. In particular, the stability of the system under varying power output conditions and the long-term reliability of the fuel under the most adverse thermal/hydraulic conditions are presented.Copyright © 2011 by ASME

Temperature of pressurized water reactor (PWR) core is a key parameter used widely for judging the initiation of emergency operating procedures and severe accident management. Since direct measurement of the fuel cladding surface temperature using thermocouples is not practicable currently, the coolant temperature at the core exit locations is monitored instead. Several experimental researches showed that the CET rise during a loss of coolant accident (LOCA) and its magnitudes were always lower than the actual fuel rod cladding temperature at the same time. In this regard, a theoretical analysis of the transient heat transfer of coolant flow in a PWR core is needed to confirm the findings from the previous experimental works. This paper addresses numerical simulation of the transient boiling-induced multiphase flow through a simplified PWR core model during a LOCA by a commercial computational fluid dynamics (CFD) code. The calculated results are discussed to understand the transient heat transfer mechanism in the core and to provide useful technical information for reactor design and operation.Copyright © 2011 by ASME

Various kinds of nozzles are attached to a pressure vessel including Steam Generator (SG) in a pressurized water reactor plant. The downcomer feedwater nozzle on the upper vessel shell and the economizer feedwater nozzle in the lower vessel shell of the SG are representative nozzles which have a non axi-symmetric shape. In most cases, external loads composed with forces and moments are imposed on those nozzles during the plant operation. In order to evaluate structural integrity of junctures between the nozzles and vessels in compliance with the ASME Boiler and Pressure Vessel Code, Section III, it is essential to find the maximum stress intensity resulting from those loads. Welding Research Council (WRC) Bulletin 297 has been used to find the maximum stress intensity since it is not straightforward to calculate the stress intensity with a non axi-symmetric two dimensional finite element model. However, the compatibility of adopting WRC Bulletin 297 to nozzles which have a variety of geometries shall be considered. Moreover, the applicability of the stress intensity resulting from the bulletin should be into consideration when interested lines where stress intensity linearization is to be performed are not exactly consistent with the line defined in the Bulletin. In this study, the nozzles in cylindrical vessel shells are developed as three dimensional finite element models, which are loaded with unit forces and moments. The stress intensities from finite element models are investigated through a comparison of WRC Bulletin 297. In addition, a methodology to apply the stress intensity results from WRC 297 to different lines is proposed.Copyright © 2011 by ASME

This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle; 3. PWR inlet nozzle; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: • To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; • To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; • To assess the significance of attached piping loads on the stresses in the nozzle corner region; and • To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.Copyright © 2011 by ASME

Flow-accelerated corrosion (FAC) in a pressurized water reactor (PWR) feedwater piping system is considered the leading degradation process that has been blamed for a number of accidents and incidents. One of the challenging issues in estimating FAC in piping is the hydrazine (N2 H4 ) oxygen (O2 ) reaction rate and its subsequent impact on the wall mass transfer. N2 H4 is injected in the system to maintain a stable pH in the water and also to act as an O2 scavenger. Previous research results indicate a varying degree of FAC rate dependence on the presence of N2 H4 in the system. The N2 H4 -O2 reaction is also a complex function of temperature and pipe material. The present paper presents a two-part analysis that uses computational fluid dynamics (CFD) tools to investigate the N2 H4 -O2 reaction and its subsequent impact on mass transfer. In the first part of the analysis, chemistry and flow of water with dissolved O2 and N2 H4 is simulated to assess different reaction mechanisms available in the literature. Results obtained from this study are compared with available experimental data for benchmarking. The numerical results were able to capture the general pattern of reaction rate as a function of temperature. Numerical simulations were also carried out to accommodate the surface effects on the reaction, but results indicate that such accommodations will yield useful results only if the geometrical extent of the near-wall-zone, where surface effects are prevalent, is well known. In the second part of the analysis, numerical simulations were carried out for a U-bend pipe. A number of restrictive assumptions were made to assess the dependence of O2 mass-transfer rate on N2 H4 -O2 reaction. Hydrodynamic results show the secondary flow pattern within the bend section. Results are also presented for the Sherwood number ratio at the pipe wall with and without reaction. Results indicate that the N2 H4 -O2 reaction decreases the O2 flow rate toward the wall. This calculation also shows that secondary flow in the bend affects the wall mass transfer pattern.Copyright © 2011 by ASME

The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI weld residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material. This paper describes the Phase 1 program, which comprised an initial period of learning and research for both FEA methods and measurement techniques using simple welded specimens. The Phase 1 specimens include a number of plate and cylinder geometries, each designed to provide a controlled configuration for maximum repeatability of measurements and modeling. A spectrum of surface and through-wall residual stress measurement techniques have been explored using the Phase 1 specimens, including incremental hole drilling, ring-core, and x-ray diffraction for surface stresses and neutron diffraction, deep-hole drilling, and contour method for through-wall stresses. The measured residual stresses are compared to the predicted stress results from a number of researchers employing a variety of modeling techniques. Comparisons between the various measurement techniques and among the modeling results have allowed for greater insight into the impact of various parameters on predicted versus measured residual stress. This paper will also discuss the technical challenges and lessons learned as part of the DM weld materials residual stress measurements.Copyright © 2011 by ASME

The U.S. Nuclear Regulatory Commission staff plans to apply a more integrated, graded approach to the review of small modular reactor (SMR) pre-application activities and design applications. The concept is to improve the efficiency and effectiveness of the reviews by focusing on safety significant structures, systems, and components (SSCs). The unique design features associated with SMRs and knowledge gained reviewing other passive reactor designs present opportunities to risk-inform the SMR design certification process to a greater extent than previously employed. The review process can be modified for SMR applications by considering the aggregate of regulatory controls pertaining to SSCs as part of the review and determining those regulatory controls which may supplement or replace, as appropriate, part of the technical or engineering analysis and evaluation. Risk insights acquired from staff reviews of passive LWR designs (i.e. AP1000, ESBWR) can also be incorporated into the review process. Further, risk insights associated with integral pressurized water reactor (iPWR) design features (i.e. Underground facilities impact on turbine missiles review) can be incorporated into the review process.

The Westinghouse Small Modular Reactor (SMR) incorporates an integral pressurized water reactor (iPWR) design in which all components associated with the nuclear steam supply system are housed within one pressure vessel. The Westinghouse SMR design also utilizes many of the key features from the AP1000 ® plant, including passive safety systems. The Westinghouse SMR will be fueled by a derivative of the successful 17×17 Robust Fuel Assembly (RFA) product. An 89 assembly core with an active height of 8 feet will provide a 24 month operating cycle with a power output of 800 MWt. Derived from the AP1000 plant and adapted to operate inside the reactor pressure vessel, 37 control rod drive mechanisms provide reactor shutdown and reactivity control capabilities. Eight seal less pumps provide a nominal reactor coolant flow of 100,000 gallons per minute. An innovative evolution of a straight tube steam generator produces a saturated mixture that is delivered to a steam separating drum located outside of the containment vessel. The steam generator along with the integral pressurizer is attached to the reactor vessel with a single closure flange located near the center of gravity of the reactor assembly and is designed to be removed during refueling operations. Like the AP1000 plant, the Westinghouse SMR relies on the natural forces of gravity and natural circulation to provide core and containment cooling during accident conditions. The passive cooling systems provide sufficient heat removal for seven days without the need for offsite AC power sources. The Westinghouse SMR also includes traditional active components such as diesel generators and pumps; however these components are not required for the safe shutdown of the plant. At a diameter of 32 feet, approximately 25 of the Westinghouse SMR containment vessels can fit within the envelope of the AP1000 containment building. This compact containment will be completely submerged in water during power operation providing a heat sink for postulated accidents. For protection against external threats, the containment vessel and plant safety systems are located below ground level. At approximately one fifth the net electrical output of the AP1000 plant, the Westinghouse SMR is designed to address infrastructure challenges associated with replacing America’s aging fossil fuel plants by providing a safe, clean and reliable energy source. The challenges associated with economies of scale are offset with a compact and simplified plant design, rail shippable components and modular construction.Copyright © 2011 by ASME

Subcooled boiling in natural circulation has great influences on the flow and heat transfer characteristics of the medium in primary circuit loop of pressurized water reactor. Onset of nucleate boiling (ONB) is a key point in boiling heat transfer. According to analyze the existing models of ONB, the main factors that affect ONB are found out. The effects of main factors and the location of ONB are analyzed based on Yang Ruichang model in natural circulation, and found a model to calculate the location of ONB in natural circulation. A mathematical model is created using unascertained mathematics and compared with other models. The conclusion illustrates that the predicting model can describe the physical phenomena of subcooled boiling efficiently, and depicts the relation of certainty and uncertainty of influencing factors preferably by unascertained mathematics, which has particular methods to handle incomplete information, unsteady and inaccuracy numerical value, and variation quantity. The model reveals the real circumstance more accurately and plays an important role in security evaluation in nuclear power plant.Copyright © 2010 by ASME

Steam generator is one of the key equipments in the pressurized water reactor, from the performance point of view, it is necessary to apply optimization techniques to the design of the steam generator. In this work, the optimal designs of a U-tube steam generator (UTSG), taking minimization of the total volume and net weight as objective respectively, are carried out by considering thermohydraulic and geometric constraints using a complex-genetic algorithm (CGA). And the sensitivities of some parameters which influence the total volume and net weight of UTSG are also analyzed. Under the condition of constant secondary thermalhydraulic parameters of the steam generator, the optimal design indicates an obvious effect taking either the overall volume or the total weight of the steam generator as the objective. The optimization results show that the proposed optimal method is feasible and effective. And the results of optimal designs and sensitivity analysis would provide guidance in the engineering design of UTSG.Copyright © 2010 by ASME

Knowledge of the quantity and distribution of core deposits (crud) can be useful in many ways. Trending the amount of crud can show the effects of coolant chemistry changes and crud remediation efforts. The crud distribution reflects the locations of sub-cooled boiling, and indirectly, core performance. Both the quantity and distribution of crud can be useful in assessing the risk of Crud Induced Power Shifts (CIPS) and Crud Induced Local Corrosion (CILC). Measuring core crud is also essential in increasing the fundamental understanding of crud transport and the build-up of ex-core radiation fields. Several techniques have been used to measure core crud levels. Crud scraping is frequently applied to sample crud, but it cannot be applied economically over broad areas. Crud has been mapped over broad core areas by analyzing the waste stream from an ultrasonic fuel cleaner, but uncertainties in the cleaning efficiency limit the accuracy of this technique. Video crud mapping, the determination of crud coverage from video images of fuel assemblies, is an economical method that can be readily applied to peripheral rods on every fuel assembly. The mapping can be performed by computer image analysis, or by visual comparisons to standards. The two techniques are compared and the advantages and disadvantages of each are discussed. The effectiveness of video crud mapping in assessing fuel performance issues is shown in the results of several field applications.Copyright © 2010 by ASME

The paper presents results of a probabilistic leak-before-break (LBB) assessment of a ferritic main coolant line of a pressurized water reactor representative for German nuclear power plants. The analysis approach is based on the elastic-plastic fracture mechanics methodology, incorporating the failure assessment diagram to calculate the critical through-thickness crack size, as well as fatigue crack growth calculations to determine the flaw length at wall penetration. An essential part of this study is the collation and statistical treatment of material data, such as strength properties, crack resistance and fatigue crack growth curves, and their incorporation in the probabilistic assessment. The analysis yields negligible break probabilities, thus demonstrating the LBB behavior of the piping. This conclusion is validated by results of a sensitivity study with additional conservative assumptions and simplifications with respect to the initial crack size and material state. For the sake of simplicity, the blocking effect of the intact austenitic cladding on the crack extension, as well as the conditional probability of crack penetration through the wall during the service life are neglected in this investigation. Through an additional treatment of these issues a more realistic assessment is achieved, resulting in even smaller failure probabilities.Copyright © 2010 by ASME

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.Copyright © 2010 by ASME

In order to develop licensing LOCA analysis program platform for nuclear power plant (NPP), a relatively realistic technique have been adopted in this paper, which is to satisfy the requirements of conservative evaluation models (EM) in related regulation (10CFR50, Appendix K) by modifying related models or correlations of the best estimate (BE) program RELAP5/MOD3, in an effort to form a licensing LOCA analysis tool. By using this “advanced program platform plus conservative EM” technique, LOCA analysis on the Chinese 300MW NPP, of which LOCA used to be conservatively evaluated, is attempted so that further potential of margins should be achieved. By comparing with Appendix K, 10 models or correlations in RELAP5/MOD3 code are identified, which need to be modified, invoked, evaluated or verified. Modifications or control have been separately introduced on such models or correlations as fission product decay model, correlations for critical heat flux (CHF), post-CHF heat transfer correlations, logic of preventing from returning nucleate boiling and transition boiling heat transfer prior to reflood, discharge model, metal-water reaction model, emergency core cooling (ECC) bypass model and related models for pressurized water reactor (PWR) refill and reflood. With the accomplishment of modification or invoking the above-mentioned models, conservatism verifications by utilizing proper standard curve or separate test data have been carried out. The standard curve or separate effect tests data includes: (1) 1971 fission product decay standard curve; (2) related ORNL THTF test data; (3) Marviken test-22 data; (4) Cathcart oxidation test data; and (5) Westinghouse FLECHT - SEASET test data, etc. Further, for the integral effect test LOFT L2–5, calculations with both original RELAP5/MOD3 program and the program of which some related models modified in compliance with the requirement of 10CFR50 Appendix K have been made. The results indicate that, on the one hand, ability of the original BE program simulating the accident progress is identified; and on the other hand, the modified models are confirmed to be conservative. Moreover, both individual contributions and combined effects of the model modification to conservatism of the integral test analysis are observed and analyzed. To assess effects of the “advanced program platform plus conservative EM” approach on LOCA analysis, finally, calculations with the modified code for both hypothetical small break and large break LOCA of the Chinese 300MWe NPP are conducted. Compared with the calculated results of original RELAP5/MOD3 program and that delivered in Final Safety Analysis Report (FSAR) of the NPP, mutual and integrated effects with different model modifications are investigated. And finally, safety margin achieved by the present calculation is preliminarily discussed.Copyright © 2010 by ASME

One of the limiting conditions during operation of a Pressurized Water Reactor is cladding integrity in case of occurrence of any conditions I or II events. The decoupling criterion is the absence of Departure from Nucleate Boiling (DNB) during the full sequence of any of these transients. Heat transfer between the clad and the water is limited by the DNB phenomenon when local surface heat flux is greater than the so-called Critical Heat Flux (CHF). Heat production at the surface is higher than heat removal capacity by the coolant therefore a vapor blanket is formed around the clad; consequently the heat transfer will drastically drop resulting in a sudden significant increase of the local wall temperature and clad damage may appear if no corrective action is initiated. DNB can not be estimated with physical principles only. Experimental support is needed for evaluation. Occurrence of DNB is evaluated using the Departure from Nucleate Boiling Ratio (DNBR) which is a function of both core thermal hydraulic (T/H) parameters and design of the fuel assembly. Advanced fuel assemblies claim higher CHF values compared to previous designs. Along with increased DNB performances for advanced fuel assemblies, CHF correlation development and advanced methodologies enable to extend normal operating conditions of a nuclear plant. On the one hand, CHF performances really increased allow additional margin related to the loss of fuel cladding integrity whereas on the other hand optimized correlations and advanced methodologies reduce this margin. An accurate assessment of the CHF performance of the advanced fuel assemblies is therefore required. This paper will raise issues regarding the assessment of the CHF performance of new advanced fuel assemblies design. The issues will be focused on the reliability of the experimental assessment of the CHF values and the accuracy of the transposition of mock up geometries to plant core configuration (representativity of the experiments). The verification that the tests conditions (pressure, flowrate, quality, heat flux …) ensure a proper coverage of all core conditions encountered during any of the conditions I & II transients is closely linked to DNBR methods and will not be extensively covered in this paper. This paper suggests some thoughts about relevance of the demonstration carried out by vendors on these matters.

In this paper, results of experiments and numerical simulations for counter-current flow in a pressurized water reactor hot leg under reflux cooling are summarized. In the experiments, we used two types of small scale PWR hot legs. One was a 1/5th scale rectangular duct, and the other was a 1/15th scale circular pipe. Air and water were used for gas and liquid phases. The air flow rate and the supplied water flow rate were varied to observe flow pattern and measure the counter-current flow limitation (CCFL) characteristics. Flow patterns in the elbow and the inclined section were strongly affected by those in the horizontal section. In the 1/15th scale circular pipe experiments, CCFL characteristics obtained by increasing the air flow rate differed from those obtained by decreasing it. CCFL characteristics corresponded to the flow pattern transition. In the numerical simulations, we used a three-dimensional two-fluid model to evaluate the capability of predicting counter-current flow in the hot leg. Good agreements between measured and predicted flow patterns and CCFL characteristics were obtained by using an appropriate set of correlations for interfacial friction coefficient. We also carried out simulations of actual hot leg conditions to examine the effects of fluid properties and size. Predicted flow patterns and CCFL characteristics were close to those of scale model calculations. We concluded the combination of calculation model and interfacial friction coefficients used in this study can predict the counter-current flow in a hot leg.Copyright © 2010 by ASME

The SMART (System-integrated Modular Advanced ReacTor) is small sized integral type pressurized water reactor designed by KAERI (Korea Atomic Energy Research Institute), Korea. But, shape of steam generator (SG) in SMART plant differs from those in operated nuclear power plants (NPPs). Especially, SG tubes in SAMRT plant is helical type with around 600 mm of innermost diameter and thickness of 2.5 mm which is thicker than general NPPs one. For providing integrity of SG tube in SMART plant, new types of ECT method are needed because eddy current testing (ECT) is one of widely adopted method for inspection of SG tubes in NPPs. Therefore, in this study, we investigate optimal conditions or parameters for detecting and evaluating of flaws in the SG tubes in SMART plant by simulation of ECT signals with various testing condition or parameter such as frequency, coil gap and etc. From the simulated ECT signals optimal eddy current test condition or parameters are proposed.Copyright © 2010 by ASME

Utilities are looking for ways to optimize plant lifetime, and must therefore prevent stress corrosion in primary components, while combating other phenomena, such as thermal fatigue or certain metallurgical weaknesses. The replacement of sections of the main primary system is one way of solving these problems. The increase in the number of the replacement of heavy components carried out in the reactor building on specific reactor geometries has called for major technical innovations on the replacement of heavy components. For above, the IAEA published a nuclear energy series (NES) on replacement of heavy components to propose guidance and share experiences. The major and heavy components to be considered are; 1) Steam generators for pressurized water reactor plants, 2) Reactor vessel head for PWR plants, 3) Reactor internal components for boiling water reactor plants, 4) Reactor vessel internals for PWR plants, 5) Pressurizer for PWR plants, 6) Reactor coolant piping/ recirculation piping PWR, and 7) Press Tube and feed piping for pressurized heavy water reactor. This paper is focused on heavy components replacement considered strategic aspects for nuclear power plants life management.Copyright © 2010 by ASME

In the course of a small break LOCA in a Pressurized Water Reactor (PWR) the flow regime in the Reactor Cooling System (RCS) passes through a number of different phases and the filling level may decrease down to the point where the decay heat is transferred to the secondary side under Reflux-Condenser (RC) conditions. During RC, the steam formed in the core condensates in the Steam Generator (SG) U-tubes. For a limited range of break size and configuration, a continuous accumulation of condensate may cause the formation of boron-depleted slugs. If natural circulation reestablishes, as the RCS is refilled, boron-depleted slugs might be transported to the Reactor Pressure Vessel (RPV) and to the core. To draw conclusions on the risk of boron dilution processes in SB-LOCA transients, two important issues, the limitation of slug size and the onset of Natural Circulation (NC) have to be assessed on the basis of experimental data, as system Thermal-Hydraulic codes are limited in their capability to replicate the complex physical phenomena involved. The OECD PKL III tests were performed at AREVA’s PKL test facility in Erlangen, Germany, to evaluate important phases of the boron dilution transient in PWRs. Several integral and separate effect tests were conducted, addressing the inherent boron dilution issue. The PKL III integral transient test runs provide sufficient data to state major conclusions on the formation and maximum possible size of the boron-depleted slugs, their boron concentration and their transport into the RPV with the restart of NC. Some of these conclusions can be applied to reactor scale. It has to be mentioned, that even though this paper is based on PKL test results obtained within the OECD PKL project, the conclusions of this paper reflect the views of the authors and not necessarily of all the members of the OECD PKL project.Copyright © 2010 by ASME

Nearly about 100 nuclear power national standards and 340 industry standards have been made in the field of nuclear power in China, which can partly meet the need of development of the second generation pressurized water reactor nuclear power plant. The situation of nuclear power fast development in China causes more standards to be made in the field of nuclear power, furthermore, the standard system and standardization have some problems now, to make more standards scientifically and solve existing problems, a number of measures should be taken in future.Copyright © 2010 by ASME

The solution to a Markov chain modeling electric power supply to critical equipment in a typical 4-loop pressurized water reactor following a Loss of offsite power event is compared with a convolution method. The standard “convolution integral” approach is described, and an alternative methodology based on a Markov model is illustrated.Copyright © 2010 by ASME

In order to see the effect of thermal aging on environmental fatigue life of CF8M cast austenitic stainless steel (CASS), low-cycle environmental fatigue tests of thermally aged CF8M CASS at the condition of fatigue strain rate 0.04%/s were conducted at the operating condition, 15MPa, 315°C of pressurized water reactor (PWR) environment. Test results of low cycle fatigue life tests for thermally aged specimens simulating 60 operating years were compared with ones of un-aged CF8M CASS in room temperature air and PWR operating conditions to see the effect of the thermal aging on environmental fatigue life. This kind of experiment would be useful to verify the fatigue integrity of long-lived components and to predict plant safety of long term operation beyond design life because current approach of evaluating environmental fatigue is so conservative to apply it to the long-lived components in pressure boundary of nuclear power plants.Copyright © 2010 by ASME

One of more frequent events in the Pressurized Water Reactor (PWR) is Steam Generator Tube Rupture (SGTR) accident, which is among the main accidents in the field of nuclear safety. This paper studies the SGTR event in the Multi-application Integrated Pressurized Water Reactor (IPWR) using the best-estimate thermal-hydraulic code RELAP5/MOD3.4. In the reactor of IPWR, several Once-Through Steam Generator (OTSG) cassettes are used and located between the core support and the pressure vessel. The tube rupture location is on the top of the tube sheet of a steam generator. Three different tube rupture modeling methods and several different subcooled discharge coefficients in the critical flow model are considered and compared. In the safety analysis, high pressure safety injection system, core makeup system and Passive Residual Heat Removal System (PRHRS) that would affect the accident consequences are considered.Copyright © 2010 by ASME

The Materials Reliability Program (MRP) Reactor Internals Focus Group (RI-FG) developed Pressurized Water Reactor (PWR) Internals Inspection and Evaluation (I&E) Guidelines under the sponsorship of the Electric Power Research Institute (EPRI). The I&E guidelines summarized in MRP-227 [1], provide a generic basis for U.S. utilities to develop their Aging Management Program (AMP) for managing the long-term aging degradation of PWR reactor internals including the existing and extended license periods. A number of internals structural bolts in the Babcock & Wilcox (B&W) design PWRs are fabricated from high-strength alloys such as Alloy A-286 or Alloy X-750. The materials in general, and bolts in particular, are known to be susceptible to stress corrosion cracking (SCC) based on past operating experience. The Upper and Lower Core Barrel (UCB and LCB) bolts have a core support function and have been generically categorized as Primary components for inspection in the I&E Guidelines. The remaining Alloy A-286 and Alloy X-750 structural bolts are in the Expansion category. Per 10CFR54, all U.S. PWRs are required to establish a unit-specific AMP for the extended license period in accordance with the ten elements of an effective AMP outlined in the Generic Aging Lessons Learned (GALL, NUREG-1801 Rev. 01, [2]) report published by the U.S. Nuclear Regulatory Commission (NRC). The goal of this paper is to provide an overview of the work performed by AREVA NP Inc. to support the development of the MRP I&E guidelines and unit-specific AMP for UCB and LCB bolts. A review of Alloy A-286 and Alloy X-750 bolts in the B&W design PWR is provided including the degradation mechanism, operating and inspection experience, replacement, and autoclave and in-reactor test results. The latest UT inspection technique used to characterize the extent of flaws is also discussed. Acceptance criteria for evaluating degraded conditions in UCB and LCB bolts were developed in accordance with the requirements of the ASME Section III, Subsection-NG core support structures requirements. In addition to Code compliance, special limits were established to limit the change in the core support structure stiffness. The acceptance criteria enable utilities to rapidly disposition UT inspection findings during an outage within 48 hours. In order to support the objectives of an efficient AMP for the UCB and LCB bolts, three-dimensional finite element models were prepared capable of evaluating all potential failure scenarios. These models enable accurate representation of flange flexibility and redistribution of loads due to deficient bolts. Prior to an outage, hypothetical patterns of bolt failures could be evaluated to support pre-outage planning and contingency preparation. During an outage, these models are used to disposition inspection results and help operability assessment of continued operation, and re-inspection requirement to ensure continued safety and integrity of the reactor vessel internals. Based on the existing work performed, future improvement and expansion of analytical capability is outlined in the last section of this paper. In conclusion, AREVA NP Inc. has demonstrated an effective use of a multi-disciplined approach using structural analyses, operating experience, material evaluations, and non-destructive examination (NDE) to fulfill both the development and implementation of unit-specific aging management commitments as required by MRP-227 for the current and extended license periods.

The major threat that nuclear power plants (NPPs) pose to the safety of the public comes from the large amount radioactive material released during design-basis accidents (DBAs). Additionally, many aspects of Control Room Habitability, Environmental Reports, Facility Siting and Operation derive from the design analyses that incorporated the earlier accident source term and radiological consequence of NPPs. Depending on current applications, majority of Chinese NPPs adopt the method of TID-14844, which uses the whole body and thyroid dose criteria. However, alternative Source Term (AST) are commonly used in AP1000 and some LWRs (such as Beaver Valley Power Station, Units No. 1 and No. 2, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 And 2, Kewaunee Power Station and so on), so it is attempted to adopt AST in radiological consequence analysis of other nuclear power plants. By introducing and implementing the method of AST defined in RG 1.183 and using integral safety analysis code, a pressurized water reactor (PWR) of 900 MW nuclear power plant analysis model is constructed and the radiological consequence induced by Main Steam Line Break (MSLB) accident is evaluated. For DBA MSLB, the fractions of core inventory are assumed to be in the gap for various radionuclides and then the release from the fuel gap is assumed to occur instantaneously with the onset of assumed damage. According to the assumptions for evaluating the radiological consequences of PWR MSLB, dose calculation methodology is performed with total effective dose equivalent (TEDE) which is the criteria of dose evaluation. Compared with dose criteria of RG 1.183, the dose of control room, exclusion area boundary and outer boundary of low population zone are acceptable.Copyright © 2010 by ASME

A set of screening and classification method is proposed in this paper to manage the Systems, Structures and Components (SSCs) of the Pressurized Water Reactor (PWR) for the aging management. The method proposed and carried out in China Nuclear Power Plants (NPP) is based on the systematic aging evaluation method and the management method for the license renewal from International Atomic Energy Agency (IAEA) and U.S. Nuclear Regulatory Commission (NRC), whose successful management methods are discussed in this paper. Considering the importance of the classification management of the SSCs, the classification method is investigated from the aspects of the safety, reliability and replacement of the SSCs with the Nuclear Power Plant operation experience feed-back. With the classification management method, the SSCs are managed economically and effectively for the Nuclear Power Plants in China. The method also makes great foundation for the aging management and residual life evaluation of the PWR in China.

During typical nuclear power plant refueling activities for a pressurized water reactor (PWR), the reactor vessel closure head assembly must be removed from the reactor vessel (RV), transported for storage, and returned to the RV after refueling. This is categorized as a critical heavy load lift in NUREG-0612 [1] because a drop accident could result in damage to the components required to cool the fuel in the RV core. In order to mitigate the potentially severe consequences of a closure head drop, the United States Nuclear Regulatory Commission (USNRC) has mandated that nuclear power plants upgrade to a single failure-proof crane, show single failure-proof crane equivalence, or perform a head drop analysis to demonstrate that the core remains covered with coolant and sufficient cooling is available after the head drop accident. The primary coolant-retaining components associated with the RV are the inlet and outlet nozzles and the hot and cold leg main loop piping. Typical head drop analyses have considered these components to ensure that their structural integrity is maintained. One coolant-retaining component that has not been included in head drop evaluations on a consistent basis is the bottom-mounted instrumentation (BMI) system. In a typical Westinghouse PWR, 50 to 60 BMI nozzles are connected through the bottom hemisphere of the RV to one-inch diameter guide tubes which run under the vessel to a seal table above. Failure of the BMI system has the potential to adversely affect core coolability, especially if multiple failures are postulated within the system. A study was performed to compare static and dynamic methods of analyzing the effects of a head drop accident on the structural integrity of the BMI system. This paper presents the results of that study and assesses the adequacy of each method. Acceptability of the BMI system pressure boundary is based on the Nuclear Energy Institute Initiative (NEI 08–05 [2]) criteria for coolant-retaining components, which are based on Section III, Appendix F of the ASME Code [3].Copyright © 2010 by ASME

Leak-Before-Break (LBB) is employed in design of nuclear power reactor piping to eliminate consideration of the dynamic effects of pipe rupture from the plant design basis for the affected piping system. LBB cannot be applied if environmental conditions that could lead to degradation by stress corrosion cracking exists. For Alloy 600/82/182 dissimilar metal welds (DMW) in pressurized water reactor plants, primary water stress corrosion cracking (PWSCC) is found to be active. Application of weld overlay (WOL) of non-susceptible Alloy 690/52/152 material has been shown to mitigate PWSCC growth in DMW. Therefore, LBB can be considered for a DMW with Alloy 690/52/152 overlay. However, WOL sizing design postulates a complex crack which is through wall in the overlay material and part through or full circumferential in the DMW base material. This significantly reduces the critical flaw size and in turn the maximum allowable flaw size for leak rate. The current industry practice conservatively ignores the full circumferential crack in the original pipe material and assumes a through wall crack along the entire pipe thickness. This assumptions leads to significantly reduced leakage due to smaller crack opening. The problem becomes more critical with small diameter pipes. The current work calculates the crack opening displacements (CODs) for a pipe with complex crack. Since it is a function of several geometry and materials parameters, response functions are generated to calculate CODs.Copyright © 2010 by ASME

The poolside examination is demanded essentially for the development of an advanced fuel, because it provides data about in-reactor fuel behaviors as well as product performance. Therefore, all the results of examinations are transmitted back to design groups, and evaluated with the design models and codes applied new fuel development. In general, PSE is performed a variety of examinations in assembly and single rod state, such as visual inspection, fuel assembly length growth, bowing and twist, rod-to-rod spacing, spacer grid width, fuel rod diameter, and fuel rod oxide thickness at the end of each cycle of irradiation during plant outage periods. After the completion of fuel life, selected rods are extracted from the fuel assembly for individual fuel rod measurement. In this paper, the techniques of PSE applied an advanced pressurized water reactor fuel, PLUS7™ for Optimized Power Reactors (OPR 1000s) in Korea are described.Copyright © 2010 by ASME

The Steam Generator Asset Management Program (SGAMP) is a long term program designed to maximize the performance and reliability of the steam generators. The SGAMP focuses on plant specific conditions and hence is applicable to the original or the replacement steam generators. It is recommended that the utility and the vendor form a joint steam generator management team (SGMT) to develop, monitor and implement a long-term plan to address steam generator operation, maintenance and life extension goals. The SGMT will consist of representatives from operations, chemistry, maintenance and engineering functions and will be responsible for making decisions related to the steam generators. The charter of the SGMT is to develop a steam generator strategic plan that will cost-effectively manage steam generator options. The strategic plan is consistent with the Steam Generator Program Guidelines (NEI 97-06 in the United States). The strategic plan is a living document and is revised periodically to incorporate inspection results, new technology developments, lessons learned and industry experience. Cost-benefit analyses of strategies may be performed to prolong steam generator operability through steam generator performance modeling (tube degradation, fouling, etc.), diagnostic tools, regulatory strategy, condition monitoring and operational assessment strategy, and maintenance strategy. The SGMT will provide input regarding potential maintenance of the steam generators with schedule and cost impacts for each outage. It will also recommend engineering evaluations to be performed in support of program goals and will develop short- and long-term recommendations. These recommendations will address action plans, performance measures and results. Secondary side inspection and cleaning strategy should be developed (techniques and frequency) to maximize performance cost-effectively. This paper is based on Westinghouse experience gained by working with several pressurized water reactor (PWR) plant operators in the United States (US).Copyright © 2010 by ASME

SKN(Shin Kori Nuclear Power Plant) Unit 3, the first unit of APR1400 reactor in Republic of Korea, which is under construction, is scheduled to realize 18-month cycle operation strategy from its initial core. SKN Unit 3 is a pressurized water reactor with a thermal output of 4000 MWth having 3-batch 241 fuel assemblies. Its target cycle length will be realized as long term(about 455 EFPD) directly from initial core, although most of previous other plants had been started according to their annual or 15-month cycle operation schedule at their initial core and gone to 18-month after 3rd ∼ 4th cycle. SKN Unit 3 has also been designed by meeting the following requirements as well as the above cycle length requirement; 1) peaking factor (Fxy) ≤ 1.60, 2) Negative MTC(Moderator Temperature Coefficient), 3) thermal margin ≥ 10%, 4) sufficient shutdown margin. 5) Xe stability, 6) CEA(Control Element Assembly) Re-arrangement, etc,. Higher enriched assemblies used to accomplish long term cycle at the initial core make higher radial and axial peaking factor than those of OPR1000. APR1400 introduces axial cutback region in a gadolinia fuel rod(6″ in each of top and bottom) and axial blanket region in a normal fuel rod(6″ in each of top and bottom) with low enriched uranium to enhance the neutron economy. It is also expected for these optimized cutback and blanket length to provide an advantage to reduce or avoid xenon oscillation phenomenon in power ascension test(PAT). In order to enhance core operability (RPCS requirement, Xe oscillation control and performance related design requirements, etc,.) as well as meeting shutdown margin requirement, CEA re-arrangement were also made. In this paper, APR1400 core design experience including all other design endeavors will be introduced.

Primary Water Stress Corrosion Cracking (PWSCC) has been observed in pressurized water reactor (PWR) coolant system pressure boundary components. Recently this cracking has been observed in Alloy 82/182 butt welds. Various repair and mitigation schemes have been proposed or employed to address this cracking. Case N-766 was developed as an alternative method for PWSCC mitigation. It is especially useful for applications in PWSCC susceptible regions where accessibility to the outer surfaces for performing other PWSCC mitigation techniques are difficult or impractical. The method in Case N-766 involves isolating the PWSCC susceptible material from its environment, in this case PWR reactor coolant fluid, thereby eliminating one of the three (3) simultaneous conditions that must exist for PWSCC to occur. Inlays provide for mitigation of flaws or repair of existing flaws and also permit maintaining the inside surface contour essentially in its original configuration, without flow path restriction. Onlays, which do not require an excavation into the pipe ID, permit maintaining the inside surface contour essentially in its original configuration, or permitting weld buildup on the inside surface. This paper presents the technical basis for the design, fabrication, and inspection requirements for inlays and onlays.Copyright © 2010 by ASME

An investigation into the increase in Plant Protection System (PPS) alarms at a three-unit US Pressurized Water Reactor (PWR) plant has determined that the alarms are the result, in part, of a hydraulic instability that has developed within the Reactor Coolant System (RCS) following the replacement of the steam generators in all three units of the Palo Verde Nuclear Generating Station (PVNGS). An experimental effort has been established by Arizona Public Service Company and Arizona State University in an attempt to determine the cause of these instabilities. Preliminary investigations have determined that the time scale of these instabilities is consistent with larger scale transient flow processes of the reactor vessel. Accordingly, the flow characteristics were assessed and localized flow measurements made using a one-fifth scale physical model of the upper plenum region of the reactor core of the Combustion Engineering System 80 reactor vessel to verify the postulation that large vortex structures referred to as “precessing” vortices [Ref. 1] affect the core exit flow conditions resulting in the noted flow instabilities. The physical model investigation was complemented by numerical analysis based on a Computational Fluid Dynamics (CFD) code performed for the same geometry. Benchmarking of the CFD model by the scaled physical model is intended to provide increased confidence in the CFD code. If verified, the CFD code may be modified so as to establish corrective actions for this condition, where physical modeling would probably be time consuming and cost prohibitive. The initial results for the physical and computational models demonstrate very good agreement between the measured and calculated flows in the upper-plenum region. The results of the complementary experimental and analytic evaluations do not support the presence of any large scale vortices of appropriate space scales that could affect flow conditions within the upper-plenum region. The elimination of the reactor vessel as the source of the instabilities suggests that the replacement steam generators may be the root cause of the flow instabilities. There is a possibility, however, that frequencies pertinent to vortices may be triggering mechanisms for flow instabilities in the entire system.Copyright © 2010 by ASME

The thermal stratification is generic to un-isolable piping system in pressurized water reactor (PWR), which is not considered in original design. This effect can finally threaten the integrity of the piping system. The pressurizer surge line is affected by thermal stratification during reactor heat up and shutdown processes particularly. It’s important to investigate and mitigate thermal stratification in the surge line to meet safety requirements. In this paper, the thermal hydraulic analysis and the mechanical analysis using numerical simulation method were conducted to evaluate the impact of thermal stratification in the surge line during reactor heat up processes. The results give the three-dimensional temperature distribution and stress status of the surge line. It indicates that thermal stratification can cause stress concentration in the displacement restricts. The mitigation of thermal stratification was also studied. The numerical simulation of pressurizer surge line can put forward thermal stratification analysis.Copyright © 2010 by ASME

After two decades of operation, austenitic stainless steel Control Element Drive Mechanism (CEDM) seal housings at a Pressurized Water Reactor (PWR) nuclear plant experienced Transgranular Stress Corrosion Cracking (TGSCC). In order to prevent the same cracking from occurring at the Fort Calhoun Nuclear Plant, a preventative program was initiated in 1999. All 37 CEDM seal housings have been inspected by using WesDyne Intraspect pancake and plus point eddy current probes. Examination of the eddy current data found that TGSCC was associated with localized areas of higher permeability (confirmed with a magnetometer). In order to quantitatively analyze the data, the normalized value from signal amplitude was defined as the arithmetic ratio between the absolute measurement of local permeability value (amplitude) and the eddy current signal value (amplitude) for the calibration standard axial notch. The data showed that in failed seal housings the normalized amplitudes were about three times greater than in non-cracked housings. Higher permeabilities were associated with cracked locations. The eddy current methodology therefore provides an empirical criterion to monitor when locally higher surface material permeability changes occur in order to determine the onset of TGSCC.Copyright © 2010 by ASME

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.Copyright © 2010 by ASME

The capacity of the electrical power system in Egypt will increase rapidly in the coming twenty years. In year 2018, nuclear power generation will be connecting to the Egyptian electrical grid. Consequently, the interaction of nuclear power plants and other systems becomes a very important issue, and a detailed nuclear power model for the medium-term and long-term power system stability should be developed. However, there is no nuclear unit model that can describe the detailed characteristics of the nuclear unit in the available commercial power system simulation software. In this paper, a detailed pressurized water reactor (PWR) nuclear unit model for medium-term and long-term power system transient stability is proposed. The model is implemented by a user defined program in PSS/E through PSS/E Matlab Simulink Interface. This model can be used to analyze the interaction of nuclear power plants and other power systems. The simulation results show that the proposed model is valid.

At present, there are a number of Generation-IV nuclear reactor concepts under development worldwide, and the SuperCritical Water-cooled nuclear Reactor (SCWR) type is one of them. The main objective of developing SCWRs is to: 1) Increase the thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs. SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25 MPa and outlet temperatures up to 625°C). This paper presents a SCWR single-reheat indirect cycle concept with intermediate heat exchangers. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, heat exchangers separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in the reactor building. The nuclear activities stay within the reactor building, and there is a reduced possibility for radioactive contamination of equipment in the turbine building. As SCW NPPs will have much higher operating thermal hydraulic parameters this paper analyzes the technical challenges and higher costs typically associated with heat exchangers. The double-pipe heat exchanger is analyzed in depth to determine the heat-transfer surface area, number of units and physical dimensions of the heat exchanger. This study will help to determine whether the advantages of the indirect cycle justify implementation of heat exchangers at a SCW NPP.Copyright © 2010 by ASME

Flaw indications have been found in some dissimilar metal nozzle to stainless steel piping welds in pressurized water reactors (PWR) throughout the world. The nozzle welds usually involve welding ferritic (often A508) nozzles to 304/316 stainless steel pipe using Alloy 182/82 weld metal. Due to an unexpected aging issue with the weld metal, the weld becomes susceptible to a form of corrosion cracking referred to as primary water stress corrosion cracking (PWSCC). It can occur if the temperature is high enough (usually >300C) and the water chemistry in the PWR is typical of operating plants. This paper represents one of a series of papers which examine the propensity for cracking in a particular operating PWR in the UK. This paper represents an examination of the weld residual stress distributions which occur in four different size nozzles in the plant. Companion papers in this conference examine crack growth and PWSCC mitigation efforts related to this plant. British Energy (BE) has developed a work program to assess the possible impact of PWSCC on dissimilar metal welds in the primary circuit of the Sizewell ‘B’ pressurized water reactor. This effort has included the design and manufacture of representative PWR safety/relief valve nozzle welds both with and without a full structural weld overlay, multiple residual stress measurements on both mock-ups using the deep hole and incremental deep hole methods, and a number of finite element weld residual stress simulations of both the mock-ups and equivalent plant welds. This work is summarized in companion papers [1–3]. Here, the detailed weld residual stress predictions for these nozzles are summarized. The weld residual stresses in a PWR spray nozzle, safety/relief nozzle, surge nozzle, and finally a steam generator hot-leg nozzle are predicted here using an axis-symmetric computational weld solution process. The residual stresses are documented and these feed into a natural crack growth analysis provided in a companion PVP 2010-25162 paper [1]. The solutions are made using several different constitutive models: kinematic hardening, isotropic hardening, and a mixed hardening model. Discussion will be provided as to the appropriateness of the constitutive model for multi-pass DM weld modeling. In addition, the effect of including or neglecting the post-weld heat treatment process, which typically occurs after the buttering process in a DM weld, is presented. During operation the DM welds in a PWR experience temperatures in excess of 300°C. The coefficient of thermal expansion (CTE) mismatch between the three materials, particularly the higher CTE in the stainless steel, affects the stresses at operating temperature. The K-weld geometry used in the steam generator nozzles in this plant combines with CTE mis-match effects to result in service stresses somewhat different from V-weld groove cases.Copyright © 2010 by ASME

The Fuel Test Loop (FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR (Pressurized Water Reactor) and CANDU (CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI (Korea Atomic Energy Research Institute). It consists of In-Pile Section (IPS) and Out-of Pile System (OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA (IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to maintain the functionality of the reactor coolant pressure boundary. The functional test and verification of the IVA were done through pressure drop, vibration, hydraulic and helium leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by a local technique and has finally been tested under high temperature and high pressure. The IVA and piping did not experience leakage, as the piping, flanges, and assemblies have been fully checked. Good data was obtained during the three cycle test which included a pressure test, pressure and temperature cycling, and constant temperature. A new concept of IVA, as easy attaching and separating test fuel at IVA will be designed thereafter.Copyright © 2010 by ASME

In an attempt to develop fatigue monitoring system, two improved fatigue evaluation schemes have been proposed to monitor fatigue degradation in major components and piping of the pressurized water reactor. Proposed methods are both aimed to obtain realistic fatigue usage factors for given plant transients. Developed schemes utilize plant operating signals such as coolant temperature, pressure and flow rate. Finite element method and an improved Green’s function approach were used to calculate stresses and fatigue usage. Case studies were performed to validate effectiveness of each proposed scheme. It has been confirmed that proposed schemes can effectively reduce excessive conservatism in estimating fatigue usage and improve accuracy in stress calculation.Copyright © 2010 by ASME

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.Copyright © 2010 by ASME

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.Copyright © 2010 by ASME