bluuebunny/crossref_metadata_2025 huggingface dataset:
Detection and Sizing of Defects in Structural Components of a Nuclear Power Plant by ECT
Authors: Chen Zhenmao , Miya Kenzo
Published: November 2008

In this paper, progress of ECT technique for inspection of stress corrosion cracks in a structural component of a nuclear power plant is reported. Access and scanning vehicle (robot), advanced probes for SG tube inspection, development and evaluation of new probes for welding joint, and ECT based crack sizing technique are described respectively. Based on these new techniques, it is clarified that ECT can play as a supplement of UT for the welding zone inspection. It is also proved in this work that new ECT sensors are efficient even for a stainless plate as thick as 15mm.
A New Concept of the 4S Reactor and Thermal Hydraulic Characteristics
Authors: Nishi Yoshihisa , Ueda Nobuyuki , Kinoshita Izumi , Koga Tomonari , Nishimura Satoshi , Yokoyama Tsugio , Maruyama Shigeki , Kimura Kimitaka , Kasai Shigeo
Published: November 2008

CRIEPI (Central Research Institute of Electric Power Industry) has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application to dispersed energy supply and multipurpose use, with Toshiba Corporation [1,2,3,4]. Electrical output of the 4S reactor is from 10MW to 50MW, and burn-up reactivity loss is regulated by neutron reflectors. The reflector that surrounds the core is gradually lifted up to control the reactivity according to core burn-up. 30year core lifetime without refueling can be achieved with the 10MW 4S (4S-10M) reactor. All temperature feedback reactivity coefficients, including coolant void reactivity, of the 4S-10M are negative during the 30year lifetime. A neutron absorption rod is set at the center of the reactor core with the ultimate shutdown rod. The neutron absorption rod used during the former 14 years is moved to the upper part of the reactor core, and the operation is continued through the latter 16 years. The pressure loss of the reactor core is lower than 2kg/cm2 to enable effective utilization of the natural circulation force, and the average burn-up rate is 76GWD/t. To suppress the influence of the scale disadvantage, loop-type reactor design is one of the candidates for the 4S-10M. The size of the reactor vessel can be miniaturized by adopting the loop type design (4S-10ML). In the 4S-10ML design, integrated equipment which includes primary and secondary electromagnetic pumps (EMPs), an intermediate heat exchanger (IHX) and a steam generator (SG) is adopted and collocated by the reactor vessel. The decay heat removal systems of 4S-10ML consist of the reactor vessel air cooling system (RVACS) and SGACS (a similar system to the RVACS, with air cooling of the outside of the integrated equipment vessel). They are completely passive systems. To decrease the construction cost of the reactor building, a step mat structure and the horizontal aseismic structure are adopted. 4S-10ML has unique features in the cooling systems such as integrated equipment and two separate passive decay heat removal systems which operate at the same time. To evaluate the design feasibility, the transition analyses were executed by the CERES code developed by CRIEPI [5]. In this paper, the design concept of 4S-10ML, and the results of the plant transition analyses are described.
The OECD Pipe Failure Data Exchange Project: Data Validation on Canadian Plants
Authors: Colligan Alexandre , Lojk Robert , Riznic Jovica
Published: November 2008

The Nuclear Energy Agency (NEA) of the Organization for Economic Co-Operation and Development (OECD) has initiated a project to establish an international pipe failure data collection and exchange program. This OECD Pipe Failure Data Exchange (OPDE) Project has been established to encourage multilateral co-operation in the collection and analysis of data relating to pipe failure events in commercial nuclear power plants. This paper presents a brief description of the ODPE project objectives and work scope, as well as the Canadian contribution on data validation with respect to development and application of the pipe failure data collection on which OPDE is based.
Regulatory Issues of Digital I&C System in Lungmen Project
Authors: Chuang C. F. , Chou H. P.
Published: November 2008

The Lungmen Nuclear Power Station (LNPS) is currently under construction in Taiwan, which consists of two advanced boiling water reactor (ABWR) units. The instrumentation and control (I&C) systems of the LNPS are based on the state-of-the-art modernized fully integrated digital design. This paper presents regulatory overviews, regulatory requirements, current major regulatory issues, as well as the areas of regulatory concerns and the lessons learned on the digital I&C systems in the Lungmen Project.
Experimental Studies on Pb-Bi Flow Technology and Steel Compatibility With Pb-Bi
Authors: Kondo Masatoshi , Takahashi Minoru , Yoshida Susumu , Sawada Naoki , Yamada Akira , Hata Koji
Published: November 2008

For the development of the Pb-Bi cooled FBR and the ADS with Pb-Bi target, the compatibility of steels for core, structural and window materials with high temperature Pb-Bi is one of the critical issues. The effect of corrosion on the mechanical strength of steels should be also concerned. An oxide layer which is formed and self-healed on the steel surface in Pb-Bi is expected to improve the compatibility if oxide potential in Pb-Bi is controlled and monitored adequately to form stable oxide layer. Therefore, monitoring technology of oxygen concentration in Pb-Bi is required. In the present study, a performance test of oxygen sensor, a steel corrosion test and a steel mechanical strength test, or a pipe rupture test, were performed as follows: (1) Test of oxygen sensor: For the monitor of the oxygen potential in Pb-Bi, a thermal stress proof type oxygen sensor made of electrolyte conductor (MgO-ZrO2 and Y2O3-ZrO2) with the reference fluid of oxygen saturated bismuth was developed, and the performance test was conducted using the corrosion test loop. The performance was stable and reliable in the 1000-hour operation. The electromotive forces (EMF) of the sensor cells of MgO-ZrO2 and Y2O3-ZrO2 were nearly the same as each other, and they were not destructed during the 1000-operation. (2) Steel corrosion test: High Cr steels including heat resisting steels were exposed to a liquid Pb-Bi flow at the temperature of 550°C, the velocity of 1m/s, the oxygen concentration of 1.7×10−8wt% and the temperature difference of 150°C for 1000 and 500 hours. It was found that weight losses were lower in general in the steels with higher Cr content. The steels with high Cr, Si and Al formed thin oxide layers and exhibited better compatibility with Pb-Bi. (3) Steel mechanical strength test (pipe rupture test): Metallurgical analysis for ruptured pipe made of SS-316 was performed. The pipe had experienced the exposure to Pb-Bi at 400°C for 3440 hours, at 350°C for 4 hours, at 300°C for 50 hours, and at 250°C for 622 hours. Pipe rupture occurred possibly due to thermal expansion of Pb-Bi at heat-up processes. The results of the analysis indicated that Pb-Bi penetration to steel matrix occurred more seriously near the ruptured part than the other part of the pipe. The analytical result suggested that a brittle fracture might occur in the inner part of the ruptured pipe wall by liquid metal embrittlement because of Pb-Bi penetration, whereas dimples observed suggested that ductile fracture might occur in the outer part of the ruptured pipe wall.
Near-Term Nuclear as the Nemesis of the Age of Pollution: Re-Engineering the Planet
Authors: Miller Alistair I. , Duffey Romney B.
Published: November 2008

The world’s carefree Age of Pollution is ending and energy technologies are scrambling for schemes that either capture their wastes or claim corresponding offsets with reductions elsewhere. Not before time. But the transition offers boundless opportunities for political and subsidies, fudges, and deceptions. Now is the time for nuclear fission, as the only technology already providing large-scale, pollution-free energy supplies, to claim proper recognition for its demonstrated pollution-free pedigree. Our industry knows that nuclear fission really is an intrinsically superior solution over those offered by the latter-day converts to sequestration and environmental cleanliness. Our singular strength is the availability of our technology. But we must inspire society with nuclear’s capacity to provide a comprehensive and affordable route to secure energy supplies with vigorous growth in its deployment, starting immediately. We must include hydrogen fuelling for vehicles to make our solution comprehensive. This paper outlines the economics. We could, however, so easily dissipate the advantage of nuclear’s availability. Programs to evolve advanced reactors should be a natural adjunct to a vigorous near-term program but it would be folly to compete for leadership in energy supply based on the promise of reactor designs that cannot be deployed until after 2020. What is needed is a transition strategy. Several countries with emerging economies are showing near-term leadership in commitment of new reactors but the developed world also needs much expanded deployment of nuclear. Without that, the transition to advanced designs after 2020 could be jeopardized.
Experiences Gained in Implementing a Broad-Based Risk-Informed Application
Authors: Schinzel Glen E.
Published: November 2008

The South Texas Project was granted a first-of-kind exemption from special treatment requirements contained in 10CFR Parts 21, 50, and 100 in August 2001. Since that time, South Texas has pursued a cautious, deliberate approach to implement these risk-informed exemption allowances. Over the past two years, South Texas has gained a unique insight into the challenges and benefits that exist in pursuing a broad-based risk-informed application. The American nuclear industry is currently pursuing similar capabilities through proposed rule 10CFR 50.69 which is scheduled for NRC final review and approval in the July, 2004 timeframe. This proposed rule closely resembles the approach taken by South Texas in the exemption process and the allowances granted. For nuclear utilities that wish to pursue a similar broadbased risk-informed application, a well-conceived strategic approach is needed to prioritize the implementation activities as well as engage stakeholders in the implementation process. Cultural and communication challenges exist which must be addressed and effectively overcome. The goal of this paper is to communicate these challenges to the attendees, inform attendees of the safety and economic benefits to be recognized through this risk-informed approach, and to provide insight into continuing application opportunities that were not readily apparent when the broad-based exemption was originally conceived. This paper and presentation will be beneficial for both domestic and international attendees, as well as for personnel with utility or regulatory backgrounds.
Development of the Methodology of the Safety Analysis Performed by the Coupled KIKO3D/ATHLET Code System in VVER-440 Type NPP
Authors: Hegyi Gy. , Hordo´sy G. , Keresztu´ri A. , Mara´czy Cs. , Panka I. , Telbisz M. , Trosztel I.
Published: November 2008

In the deterministic safety analysis codes are required in order to provide evaluations of potential nuclear plant accidents. In the fields of the core transient behaviour, the computer codes have achieved a high degree of realistic modelling. Nevertheless, some further tools for the investigations of the wide range of physical phenomena in the whole plant transient, such as modeling the ex-core detector signals and the malfunctioning of the emergency control system are unavoidable, too. The programs and methods used in KFKI-AEKI for safety analysis of VVER-440 NPP are presented. The accident analysis methodology for a boron dilution scenario, in which an inactive coolant loop is started, is shown.
Lead-to-CO2 Heat Exchangers for Coupling of the STAR-LM LFR to a Supercritical Carbon Dioxide Brayton Cycle Power Converter
Authors: Moisseytsev Anton V. , Sienicki James J. , Wade David C.
Published: November 2008

Recent development of the Secure Transportable Autonomous Reactor-Liquid Metal (STAR-LM) lead-cooled natural circulation fast reactor (LFR) has been directed at coupling to an advanced power conversion system that utilizes a gas turbine Brayton cycle with supercritical carbon dioxide (S-CO2) as the working fluid. A key ingredient in achieving a coupled plant having a high efficiency are the modular lead-to-CO2 heat exchangers that must fit within the available volume inside the reactor vessel and must heat the S-CO2 to a high temperature. Thermal hydraulic performance and feasibility of seven different heat exchanger concepts has been investigated with respect to the achievement of a suitably high Brayton cycle efficiency for the coupled LFR-S-CO2 plant. The relative merits of the different heat exchanger configurations are revealed by the analysis which provides a basis to select the most promising concepts for further development.
Direct Vessel Vertical Injection for Advanced Emergency Core Cooling System
Authors: Yoon Sang H. , Suh Kune Y. , Yu Yong H.
Published: November 2008

A direct vessel vertical injection (DVVI) system, which is an improved variation of the current direct vessel injection (DVI) system for the APR1400 (Advanced Power Reactor 1400 MWe), is being developed by PHILOSOPHIA, Inc. and the Seoul National University in collaboration with the Korea Atomic Energy Research Institute (KAERI). The DVVI aims to sizably enhance penetration of the emergency core coolant (ECC) into the downcomer. This industry-academia-laboratory trilateral project is being carried out at the THETA (Transient Hydrodynamic Engineering Test Apparatus) apparatus. Once the engineering design is verified and validated, the system is slated to be employed in the more advanced reactor D2R2 (Digital R&D Reactor) to cope with loss-of-coolant accidents (LOCAs). While the ECC injection through the DVI system is known to cause multidimensional thermal hydraulic phenomena such as the impingement and breakup in the reactor downcomer, the DVVI injection system can deliver more ECC into the lower plenum and the core because of the increased downward momentum. CFX, which is a multidimensional fluid dynamics code, was used to analyze the flow pattern for the water from the DVVI versus DVI nozzles and the air from the cold legs. Results indicated that the two-phase flow distributions and downward velocities between the DVVI and DVI systems were considerably different and that the ECC penetration path into the core was clearer in the DVVI system. Experiments were performed in the 1:10 scaled-down test apparatus to measure the water-air two-phase direct bypass ratio for the DVVI system, which was less than that for DVI. The direct bypass was visualized and measured. Once the nozzle manufacturing and assembly procedures are engineered, results of this work may well contribute to improving the ECC penetration capability in the advanced reactors like D2R2.
Development of the Methodology for Application of Revised Source Term to Operating Nuclear Power Plants in Korea
Authors: Park Moon Soo , Kang Chang-Sun , Moon Joo-Hyun
Published: November 2008

Considering the current trend in applying the revised source term proposed by NUREG-1465 to the nuclear power plants in the U. S., it is expected that the revised source term will be applied to the Korean operating nuclear power plants in the near future, even though the exact time can not be estimated. To meet the future technical demands, it is necessary to prepare the technical system including the related regulatory requirements in advance. In this research, therefore, it is intended to develop the methodology to apply the revised source term to operating nuclear power plants in Korea. Several principles were established to develop the application methodologies. First, it is not necessary to modify the existing regulations about source term (i.e., any back-fitting to operating nuclear plants is not necessary). Second, if the pertinent margin of safety is guaranteed, the revised source term suggested by NUREG-1465 may be useful to full application. Finally, a part of revised source term could be selected to application based on the technical feasibility. As the results of this research, several methodologies to apply the revised source term to the Korean operating nuclear power plants have been developed, which include 1) the selective (or limited) application to use only some of all the characteristics of the revised source term, such as release timing of fission products and chemical form of radio-iodine and 2) the full application to use all the characteristics of the revised source term. The developed methodologies are actually applied to Ulchin 9&4 units and their application feasibilities are reviewed. The results of this research are used as either a manual in establishing the plan and the procedure for applying the revised source term to the domestic nuclear plant from the utility’s viewpoint; or a technical basis of revising the related regulations from the regulatory body’s viewpoint.
Study on the Advanced Concept for Shortening Construction Period of ABWR Plant
Authors: Ijichi Hiroshi , Satoh Takashi , Yamane Tadashi , Fujita Tomohiro , Tsutagawa Masahiro , Mori Hiroya , Hirako Shizuka , Murayama Kouichi , Sakakibara Hisato , Orito Yoshihiro
Published: November 2008

Since the construction of a nuclear power plant (NPP) requires a very long period, we have to wait for a long time before getting return on the investment. Furthermore, it is unfavorable that the longer construction period makes the construction cost higher. In order to improve these issues, we are studying the way of shortening construction period for the ABWR. In the previous study, we introduced a concept for shortening construction period and master schedule (M/S) based on the large-scale module method. In this study, we have developed the advanced concept for shortening construction period of the ABWR plant based on the rearrangement of its general arrangement (GA) and the application of large-scale composite modules. The estimated construction period is 22.5 months from the 1st concrete work to the fuel loading (F/L).
Design Concept of Composite Module for Nuclear Power Plant Construction
Authors: Yotsuya Taihei , Murayama Kouichi , Miura Jun , Nakajima Akira , Kawahata Junichi
Published: November 2008

A composite module construction method is to be examined reflecting one of the elements of construction rationalization of a future nuclear plant planned by Hitachi. This concept is based on accomplishments and many successes achieved by Hitachi through application of the modular construction method to nuclear power plant construction over 20 years. The feature of the composite module typically includes a planned civil structure, such as a wall, a floor, and a post, representing modular components. In this way, an increased level of rationalization is expected in the conventional large-scale nuclear plants. Furthermore, the concept aiming at the modularization of all the building parts comprising medium- or small-scale reactors is also to be examined. Additional aims include improved reductions in the construction duration and rationalization through use of the composite module. On the other hand, present circumstances in nuclear plant construction are very pressing because of economic pressures. With this in mind, Hitachi is pursuing additional research into the introduction of drastic construction rationalization, such as the composite module. This concept is one of the keys to successful future plant construction, faced with such a severe situation.
Modeling of Safety/Relief Valves With Thermal-Hydraulic System Computer Codes
Authors: Delfino Claudio , Aktas Birol
Published: November 2008

A generic guideline for thermal-hydraulic (T-H) simulation of multiple bank safety relief valves (SRVs) was developed. To test the guideline, the Full Integral Simulation Test (FIST) 6PMC2 was simulated with the consolidated T-H code of USNRC, TRACE. The FIST 6PMC2 experiment simulates the response of a generic BWR/6 plant to a Main Steam-line Isolation Valve (MSIV) closure without reactor scram. During the test, the HPCS is unavailable and not used. The only inventory make-up system available is the Reactor Core Isolation Cooling (RCIC). This experiment can also be considered proto-typical of a BWR ATWS-like scenario. The simulation is largely dominated by the transient behavior of the SRVs. In this study, the experimental data was analyzed and used to check the modeling guideline for SRVs. The guideline relies on only the data available prior to an experiment or any other analysis, e.g. valve flow coefficients, inlet hydraulic diameters, etc. The study also revealed deficiencies in the “then” current valve model of the TRACE code which were subsequently corrected. The study demonstrates that the T-H models can simulate the operational behavior of SRVs very accurately while rather simple mistakes can be very damaging at the same time.
Analysis of a Convection Loop for GFR Post-LOCA Decay Heat Removal
Authors: Williams Wesley C. , Hejzlar Pavel , Saha Pradip
Published: November 2008

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA GFR. The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO2 outperforms helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops.
Sodium and Steam Generator Leak Detection for Prototype Fast Breeder Reactor (PFBR)
Authors: Vijayakumar V. , Chetal S. C. , Madhusoodanan K. , Paramasivan Pillai C. , Sakthivel M. , Seshadri Uma
Published: November 2008

Construction of Prototype Fast Breeder Reactor (PFBR) a 500 MWe pool type sodium cooled breeder reactor with MOX fuel has started at Kalpakkam. Instrumentation & Control (I&C) of PFBR is designed for safe, reliable and economic operation of the plant. Special feature of breeder reactors is sodium instrumentation. Leaks in sodium systems have the possibility of being exceptionally hazardous due to the reaction of liquid sodium with oxygen and water vapour in the air. In addition, leakage from primary systems can cause radioactive contamination. Potential regions of leakage are near welds and high stress areas. Sodium also reacts with concrete releasing hydrogen and leading to damage and loss of strength of concrete structures. Leaking sodium catches fire depending on its temperature. Sodium temperature in the plant ranges from 423 K at filling condition to 820 K at reactor nominal power operating condition. Leak detectors are provided on pipelines, tanks and other capacities. Sodium leak detection systems are designed to meet requirements of ASME section XI- division 3 which specifies that sodium leak at the rate of 100 g/h are to be detected in 20 h for air filled vaults and 250 h for inert vaults. Diverse leak detection methods are employed for active and non-active sodium equipment and pipes. For detection of water leaks into Sodium in steam generators, Hydrogen in Sodium Detectors (HSD) are used. Hydrogen in Argon Detectors (HAD) are used for sodium temperatures below 623 K as HSD is not effective below this temperature due to non-dissolution of hydrogen formed. Choice and challenges posed in implementation of above leak detection requirements are discussed in this paper.
A Study on the Hydrogen Embrittlement Mitigation of Carbon and Low Alloy Steel for Nuclear Power Plant Piping
Authors: Park Jin-Seok , Kim Jong-Sung , Kim Jung-Gu , Jin Tae-Eun
Published: November 2008

A new concept in implementing hydrogen embrittlement (HE) mitigation on nuclear piping materials was suggested by using electro-transport treatment. The electro-transport treatment can be applied to reduce HE susceptibility and thus make the material more resistant to hydrogen damage. To demonstrate the validation of the developed method, the effect of hydrogen on the tensile properties of SA-508 Grade 1A low alloy steel was investigated by slow strain rate test. Material characteristics were observed as a function of hydrogen charging condition and electro-transport treatment. Based on these experimental results, the electro-transport treatment was proposed as a potential method for mitigating the HE susceptibility to replace the conventional post-weld heat treatment.
Expansive Development of a Decommissioning Program “Recycle Simulator” in Nuclear Power Station
Authors: Nishiuchi Tatsuo , Ozaki Sachio , Hironaga Michihiko
Published: November 2008

A decommissioning program “RECYCLE SIMULATOR” should be put into practice in careful consideration of both recycle of non-radioactive wastes and reduce of radioactive wastes in the coming circulatory social system. Nevertheless current support systems for decommissioning planning mainly deal with decontamination, safety storage and dismantlement, so-called the prior part of the total decommissioning process. Authors emphasize the necessity of total planning of decommissioning including recycle or reuse of a large amount of demolition materials and are propelling the development of the multi expert system named “RECYCLE SIMULATOR”. This paper presents an algorithm of the recycling and reusing scenario of demolition materials and a summarized configuration. RECYCLE SIMULATOR for the demolished concrete was developed in 2000 and presented at 10th International Conference on Nuclear Engineering (icone10). Construction of a supporting multi expert system for the totally planning of decommissioning projects is objected by expansive development of the previous version. Main conclusions obtained from this paper are under this. (1) The previouslly developed expert system was advanced in its estimational function toward the satisfaction of decommissiong planners. (2) The applicability of the system was enlarged to all the radioactive and non-radioactive wastes, demolished metal and concrete products, in a corresponding site of decommissioning. (3) Finally decommissiong recycle simulator was completed in a harmonized unification.
Assessment of RELAP5 mod3.3 and CATHARE2 V1.5a Against a Full Scale Test of PERSEO Device
Authors: Bianchi F. , Meloni P. , Ferri R. , Achilli A.
Published: November 2008

PERSEO device was developed in the framework of a domestic research program on innovative safety systems, with the purpose to increase the reliability of passive Decay Heat Removal Systems implementing in-pool heat exchangers. The device was tested at SIET Thermal-hydraulic Research Centre by modifying the existing PANTHERS IC-PCC facility. Two types of tests were performed: integral tests and stability tests. The experimental data acquired in the test campaign allowed a validation of a RELAP5/mod 3.3 beta release and CATHARE2 V1.5a/Mod8.1 full scale model of the PERSEO device. The paper deals with the comparison between the two codes against an integral test considered representative from the point of view of the PERSEO functioning and it highlights capabilities and limits of the codes in simulating such kind of test.
Using RIAM for Optimizing Reactor Vessel Head Leak Failure Mode Maintenance Strategies
Authors: Erguina Vera , Kee Ernest J.
Published: November 2008

A cost-benefit-risk study of the best (in terms of plant valuation) long-term strategy for canopy seal weld leakage control and prevention including a span of options for reactor vessel head J-Groove weld failures is performed. The range of options goes from the “do nothing” option up to replacement of the reactor vessel head with new control rod drive mechanisms. The results are presented in form of net present values for the options cost and include uncertainties for long-term costs associated with the options activities and prediction of failure rates for control rod drive mechanisms canopy seals and J-Groove welds. Results of this study rely on catching leaks during an outage and not during operation (mid-cycle). The consequence associated with a mid-cycle leak is much higher than for one found during an outage. This tends to indicate rigorous visual inspections during outages to ensure repairs are made during refueling outages. Thus, the effect of forced outage due to mid-cycle leak detection is not included to avoid strongly biasing the results in favor of reactor vessel head replacement.
Assessing Aircraft Hazards to Nuclear Facilities
Authors: Sandquist Gary M. , Sentell D. Shannon
Published: November 2008

The methodology and selected models used for assessing the frequency of aircraft crashes to critical surface facilities are examined. The DOE Standard model is the basis for comparing other models, particularly those used to assess risk to DOE facilities. The NRC methodology is relevant for nuclear power plants. A rigorous physical model that describes the Markov chain of events and their related probabilities that lead to aircraft hazards to ground facilities is developed. The chain of events is then quantified to provide a basis for comparison and evaluation of mathematical models. It is evident that all quantitative models for aircraft crash frequency assessments are constrained by the limited statistical database available for supporting such risk assessments. Aircraft crashes to critical facilities are rare events, and this condition limits experimental data and verification efforts.
The Effect of Environmental and Mechanical Variables on Stress Corrosion Cracking of a Martensitic Stainless Steel for Transmutation Applications
Authors: Prabhakaran Ramprashad , Roy Ajit K.
Published: November 2008

Extensive work has been performed on Type 422 stainless steel (SS), to characterize environment-induced degradations in aqueous environments of different pH values at ambient and elevated temperatures. The test material was thermally treated prior to the evaluation of stress-corrosion-cracking (SCC) behavior by slow-strain-rate (SSR) and constant-load (CL) testing techniques, using smooth and notched tensile specimens. Cyclic potentiodynamic polarization (CPP) testing was performed to evaluate localized corrosion behavior using a three-electrode polarization technique in similar environments. Fractographic and metallographic evaluations of broken specimens were also performed by scanning electron microscope (SEM) and optical microscope, respectively.
Characterization of Oxide Film on the Surface of SCC in PLR Pipe by Micro Raman Spectroscopy and Its Implication to Crack Growth Characteristics at Onagawa Nuclear Power Plant
Authors: Kai Akira , Short Michael P. , Terayama Yuichiro , Tatsuki Tadashi , Watanabe Takeshi , Shoji Tetsuo
Published: November 2008

Micro Raman Spectroscopy (MRS) is known as one of the most convenient techniques for surface analysis. In this study, a MRS system was applied to analyze oxides on a stress corrosion cracking (SCC) fracture surface at the site of a nuclear power plant in Onagawa, Japan. The sample was cut from a cracked component of the Primary Loop Recirculation (PLR) Pipe of the Onagawa Nuclear Power Plant in operation for 14.6 EFPY. Scanning Electron Microscopy (SEM) and Energy Dispersive X-ray Spectroscopy (EDX) were also applied for observation and elemental analysis of the fracture surface. The oxide film on the fracture surface and the oxide particles on the film were observed with SEM. These oxides were identified as one or more of the following spinel oxides: NiFe2O4, Fe3O4, and FeCr2O4. These were present over the whole crack surface. It is worthwhile to note that α-Fe2O3 was detected near the crack mouth and Cr enrichment in the oxides was detected at the crack tip. Although α-Fe2O3 formed on the surface of austenitic stainless steel exposed to oxygenated water, since conditions in the shallow crack maintained a low pH and a low electrochemical potential, Cr enrichment and a lack of α-Fe2O3 would be expected at the crack tip. On the other hand, when the crack opening became large α-Fe2O3 was observed. Since oxides formed appeared to correspond to the size of the crack opening, oxide analyses were performed on the fracture surfaces of two different size specimens. SCC was generated in these specimens with different crack growth rates in a simulated BWR environment in the laboratory. α-Fe2O3 was observed near the crack tip in the small specimen with a slow crack growth rate. A correlation between the distribution of oxides and the size of the crack opening was extrapolated. It is likely that the existence of α-Fe2O3 near the crack tip in the plant sample is associated with either blunting or a large crack opening caused by crack growth retardation. Finally, it is assumed that the crack in the plant sample grew in the circumferential direction rather than the radial direction since α-Fe2O3 was observed closer to the tip of the central part of the crack rather than to the side edge.
An Implementation and Assessment of the Viscous Stress Model in the MARS Multi-Dimensional Component
Authors: Bae S. W. , Jeong J. J. , Hwang M. , Chung B. D.
Published: November 2008

MARS is a best-estimate thermal hydraulic systems analysis code with a multi-dimensional analysis capability. The MARS code is developed with the safety analysis as the prime application area and has a comprehensive range of models and components suitable for the simulation of nuclear reactor systems. The current developments in the nuclear industries and researches require accurate multi-dimensional analysis capabilities. However, the multi-dimensional model of MARS is based on a sub-channel formulation, and as such the geometric description is significantly limited. In order to alleviate such limitations, a new component based on the Cartesian and cylindrical coordinates, and the incorporating lateral convective terms and viscous stress terms have been added. The newly developed multi-dimensional component makes it possible to model a large bulk fluid volume in both the Cartesian and cylindrical coordinates. For the implementation of viscous shear stresses, Prandtl’s mixing length model is simply applied to obtain the turbulent viscosity. Several conceptual problems are described and solved using the new multi-dimensional component. The results were compared with the FLUENT calculations using the standard k-ε model. The comparison shows successful implementations and reasonable agreements with the FLUENT results.
Recent Improvements of Physical Models in the CATHARE Code and Their Validation
Authors: Dor I. , Geffraye G. , Lavialle G. , Mieusset T.
Published: November 2008

CATHARE is a system code developed by CEA, EDF, FRAMATOME-ANP and IRSN for Pressurized Water Reactor (PWR) safety analysis. Two-phase Flows are described using a Two-Fluid Six-equation model (1 D module) and the presence of non-condensable gases can be taken into account by one to four additive transport equations. The code allows a three-dimensional modelling of the Pressure Vessel (3D module). Successive sets of closure laws are developed in an iterative methodology of improvement. The main recent developments are presented and discussed. Film condensation of steam in presence of nitrogen and helium in a tube has been investigated in the COTURNE experiment. The liquid film heat transfer coefficient is first investigated in pure saturated steam for the wavy laminar film regime. In presence of non-condensable gases, the mass diffusion of steam in the gas mixture is modelled using a Sherwood number and a heat and mass transfer analogy. The code shows good agreement with the COTURNE tests and with other experimental data of Siddique & al and Nagasaka & al. The CATHARE capability to well represent the entrainment of liquid from the upper plenum to the steam generator inlet chamber is validated using MHYRESA experimental data. A new stratified interfacial friction based on experimental data analysis leads to satisfactory CATHARE predictions compared to MHYRESA and UPTF-TRAM data. The CATHARE 3D module capability to well calculate the refilling phase of the lower plenum considering the multi-dimensional effects is validated using UPTF tests. Two types of tests are considered to separate mechanical and condensation effects. Considering a specific entrainment rate and condensation rate for such an annular downcomer, CATHARE predictions are in good agreement with the experimental results. The BETHSY test 9.1b (ISP27) consists of 2-inch cold leg break with an ultimate procedure actuated on a cladding temperature criterion. The comparison between the calculation and the experimental data shows a rather good agreement.
Containment Design, Performance Criteria and Research Needs for Advanced Reactor Designs
Authors: Bagchi Goutam , Ali Syed , Costello James
Published: November 2008

This paper points out some important shifts in the basic expectations in the performance requirements for containment structures and discusses the areas where the containment structure design requirements and acceptance criteria can be integrated with ultimate test based insights. Although there has not been any new reactor construction in the United States for over thirty years, several designs of evolutionary and advanced reactors have already been certified. Performance requirements for containment structures under design basis and severe accident conditions and explicit consideration of seismic margins have been used in the design certification process. In the United States, the containment structure design code is the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NE-Class MC for the steel containment and Section III, Division 2 for reinforced and prestressed concrete reactor vessels and containments. This containment design code was based on the early concept of applying design basis internal pressure and associated load combinations that included the operating basis and safe shutdown earthquake ground motion. These early design criteria served the nuclear industry and the regulatory authorities in maintaining public health and safety. However, these early design criteria do not incorporate the performance criteria related to containment function in an integrated fashion. Research in large scale model testing of containment structures to failure from over pressurization and shake table testing using simulated ground motion, have produced insights related to failure modes and material behavior at failure. The results of this research provide the opportunity to integrate these observations into design and acceptance criteria. This integration process would identify “gaps” in the present knowledge and future research needs. This knowledge base is important for gleaning risk-informed insights into the design basis and severe accident behavior of containments.
Heat Transfer Characteristics of the Steam Reformer in the HTTR Hydrogen Production System
Authors: Takeda Tetsuaki
Published: November 2008

A technology development of a hydrogen production system by nuclear heat is being performed as a heat application system of a high-temperature gas-cooled reactor in the Japan Atomic Energy Research Institute. The objectives of this study are to clarify the heat transfer performance of the steam reformer in the HTTR hydrogen production system and to obtain characteristics of heat transfer and pressure drop in a channel having metallic wire inserts with high porosity. A heat transfer experiment has been performed using a horizontal circular tube. It was found that the heat transfer performance of the method using the metallic wire inserts could be further improved under the high temperature conditions.
Contact Angle of Water Droplets in a High-Temperature, High-Pressure Environment
Authors: Hayashi Tsukasa , Hazuku Tatsuya , Takamasa Tomoji , Takamori Kenrou
Published: November 2008

This paper presents an experimental study of surface wettability on a stainless plate in a high-temperature, high-pressure environment. Using a pressure vessel, we measured contact angles of water droplets at temperatures from 20 to 300°C and a constant pressure of 15 MPa, as an indicator of macroscopic surface wettability. Measured contact angles decreased with temperature below 250°C, clustering around a straight line at temperatures below 120°C and around another line in the range from 120 to 250°C. At temperatures above 250°C, on the other hand, the contact angles remained constant, independent of temperature, and contrary to the existing theoretical model, no highly hydrophilic condition or null contact angle condition was achieved.
Experimental and Analytical Study on Chemical Reaction Loss Accident With a Mock-Up Model of HTTR Hydrogen Production System
Authors: Ohashi Hirofumi , Nishihara Tetsuo , Takeda Tetsuaki , Maeda Yukimasa , Sato Hiroyuki , Inaba Yoshitomo , Hayashi Koji , Takada Shoji , Inagaki Yoshiyuki
Published: November 2008

Japan Atomic Energy Research Institute has been developing a hydrogen production system coupling with the High Temperature Engineering Test Reactor (HTTR). In the design of a HTTR hydrogen production system, a steam generator (SG) is installed at the downstream of a chemical reactor, a steam reformer (SR), as a thermal absorber to keep the secondary helium gas temperature within an allowable value at an intermediate heat exchanger inlet to prevent a reactor scram. Prior to construction of it, a mock-up model test facility, approximately 1/30-scale model of the HTTR hydrogen production system, was fabricated and simulation test on chemical reaction loss accident has been carried out to investigate the mitigation of thermal disturbance using SG. It was confirmed that fluctuation of helium gas temperature could be mitigated within a target range from −10 to 10°C at SG outlet, whereas SG inlet helium gas temperature increased 231 °C, and numerical analysis results agree well with experimental results.
Risk-Informed Pressurized Water Reactor Vessel Inspection Interval Extension
Authors: Boggess Cheryl L. , Bishop Bruce A. , Palm Nathan A. , Hedden Owen F.
Published: November 2008

The methodology discussed in this paper provides a risk informed basis for decreasing the frequency of inspection for the Pressurized Water Reactor (PWR) reactor pressure vessel (RPV). The decrease in frequency is based on extending the interval between inspections from the current interval of 10 years to 20 years. Results of pilot studies on typical designs of PWR vessels show that the change in risk associated with extending the inspection interval by more than 10 years is within the guidelines specified in U.S. Regulatory Guide 1.174 for insignificant change in risk. The current requirements for inspection of reactor vessel pressure-containing welds have been in effect since the 1989 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, supplemented by U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.150, June 1981. The manner in which these examinations are conducted has recently been augmented by Appendix VIII of Section XI, 1996 Addenda, as implemented by the NRC in amendment to 10CFR50.55a effective November 22, 1999. This paper summarizes the insignificant change in risk results for the PWR pilot-plant studies, including the effects of fatigue crack growth and in-service inspection of postulated surface-breaking flaws. These results demonstrate that the proposed RPV inspection interval extension is a viable option for the industry.
A Review of Root Causes of SCC Phenomena in BWR/RBMK: Issues, Approaches, and Proposals
Authors: Saji Genn
Published: November 2008

The author looked for potential mechanisms deemed insignificant in the early stages of SCC research and the current approach used in SCC studies were selected. The basis for those mechanisms was in non-nuclear corrosive environments as well as analyses of redox potential data measured under radiation. Three possible mechanisms were identified; ‘long cell action (which suppresses local cell cathodic activities and accelerates remote local cell anodic activities),’ ‘autocatalytic growth of cracks in crevice water chemistry,’ and ‘transpassive corrosion of stainless steel.’ The ‘long cell action’ has been identified as a key mechanism of soil corrosion which is induced by a difference in the availability of oxygen inside the soil. In this mechanism, conduction of electrons through an electric conductor over a long distance plays a significant role. This author proposes a hypothetical mechanism that states; ‘radiation-induced ‘long cell action’ causing electrolytic corrosion.’
Void Fraction Predictions in Rod Bundles at Low-Pressure Low-Flow Conditions Based on Cunningham-Yeh Model
Authors: Frepoli Cesare , Ohkawa Katsuhiro
Published: November 2008

Many experiments have been conducted in the past with full-scale rod bundles to develop void fraction correlations or interfacial drag model which can be used to predict the mixture level in a reactor core following a postulated Loss of Coolant Accident (LOCA). The Cunningham and Yeh correlation was originally developed and validated with boil-off data obtained in the 100 to 400 psia pressure range. Subsequently the validity of the correlation was successfully assessed against several other experiments. However most of the data concentrated in the intermediate to high pressure range (from 100 to 2200 psia). More recently, the development of advanced passive plant such as the AP1000, put more emphasis in the level swell behavior in the near-atmospheric pressure conditions. Following a postulated SBLOCA event for the AP1000 design, the reactor vessel is automatically depressurized to a near atmospheric condition and in the long term the core decay heat is removed by gravity driven injection flow while boiling is occurring in the core. In this paper the Cunningham-Yeh correlation was assessed against data beyond its original data base. Cunningham-Yeh model predictions were compared to several low-flow, low-pressure full-scale rod bundle experiments. Results show that the correlation performs relatively well against low pressure test data. However the Cunningham-Yeh model has the tendency to underpredict the void fraction and therefore to provide conservative results of level swell for plant safety analysis.