bluuebunny/crossref_metadata_2025 huggingface dataset:
Liquid Metal Reactor Regulatory Framework Assessment
Authors: LaChance Jeffrey L. , Duran Felicia A. , Phillips Jesse , Bari Robert , Budnitz Robert J. , Cahalan James , Flanagan George
Published: February 2010

This paper summarizes an assessment of the regulatory framework and requirements for licensing a liquid metal reactor (LMR) for use in transmuting actinides, which was performed for the U.S. Department of Energy (DOE) Advanced Fuel Cycle Initiative (AFCI). Since the LMR designs currently under consideration are sodium-cooled, the assessment identifies and discusses requirements, issues, and topics important to the licensing process in general and those specific to sodium-cooled LMRs, as well as licensing options and associated recommendations. The goal of the regulatory framework assessment was to clarify and evaluate requirements that support the development of safe and cost-effective LMR designs. The scope of the assessment included an analysis of past and present licensing practices as well as an examination of possible future regulatory activities needed to support licensing LMR designs. Because this assessment included the identification of potentially problematic areas, a review of the past LMR licensing efforts was performed. Both technical and regulatory issues were identified and recommendations were made to address important issues. A review of the current regulatory framework for licensing a commercial reactor and the associated licensing schedules was performed as part of the assessment. In addition, specific options proposed by the U.S. Nuclear Regulatory Commission (NRC) for licensing an LMR were also assessed with regard to their potential impacts on different stakeholders, which include the NRC, DOE, industry, and the public. In addition to the licensing of a commercial LMR, the assessment also identifies and evaluates licensing options for an LMR prototype. The regulatory assessment supports a conclusion that a safe, licensable LMR design is fully feasible. The knowledge applied in the LMR design will be reinforced by past experience and available technology. The licensing of an LMR is expected to be manageable, notwithstanding the uncertainties associated with regulatory, technical, and other issues. With forward-looking planning, effective management, and adequate resources, the process of obtaining a license for an LMR would be greatly facilitated.
Experimental Measurements of Eddy Current Signal From SG Tubes of Fast Breeder Reactor Covered by a Thin Sodium Layer Using a SG Mock-Up
Authors: Yamaguchi Toshihiko , Mihalache Ovidiu , Ueda Masashi , Miyahara Shinya
Published: February 2010

In Fast Breeder Reactors (FBR) which are sodium cooled, the steam generator (SG) heat exchanger tubes separate the low pressure sodium flowing in the SG vessel with the high pressure water-steam in tubes. During In-Service Inspection (ISI), sodium is first drained and then SG tubes are cooled down to the room temperature. After sodium draining, due to the high temperature (more than 500 °C), sodium adheres to SG tubes and structures around (SG support plates, welds) in a thin layer, filling eventually the gaps between SG support plates and tubes. During ISI, SG tubes are inspected for cracks and corrosions using differential eddy currents (EC) probes. Due to the high electrical conductivity of sodium adhering to the outer SG tube surface, the eddy current testing (ECT) signal modifies, in accord with sodium layer thickness or sodium deposits located on the outer SG tube surface. The sodium wetting properties depends on several factors as: material surface, temperature and sodium wetting time. The effect of sodium adhering to the outer SG tube on ECT signals were measured using a small mock-up tank (2 m high and 0.7 m in diameter) in which were introduced two SG tubes similar with the ones used in the Monju FBR (one tube is ferromagnetic and made of 2.25Cr–1Mo alloy, while the other one is made of SUS321 and is austenitic). Defects, SG support plates (on both helical and straight part of the tube) and welds were added to tubes and the ECT signal was measured before and after sodium draining. Variations in the sodium layer thickness and consequently its effect on ECT signals were measured by filling and draining the tank three times in order to recreate each time new layers of sodium. The paper describes the experimental conditions and the ECT results for both types of SG tubes by comparing the defects, SG support plates and weld signals before and after draining of sodium. Additionally, sodium structures were examined visually using a VideoScope camera, confirming the recorded ECT signals. The paper also presents details about sodium layer thickness measurements in several parts of SG tubes (near defect, SP, weld, bend, helical tube, straight tube) by scratching and collecting the sodium on a small area of 20mm×20mm. The volume of sodium drops is also estimated. The measurement results showed that there are significant differences in the sodium layer thickness depending on the SG tube material.
Electromagnetic Pulse Technology as a Means of Joining Generation IV Cladding Materials
Authors: McGinley John
Published: February 2010

The Institute for Transuranium Elements (ITU) has conducted investigations into the possibility of using Electromagnetic Pulse Technology (EMPT) in joining PM 2000 ODS and T91 ferritic-martensitic steels, using samples fabricated to simulate tube and end caps. The investigations are primarily directed towards the joining of cladding tubes to enable fuel pins to be produced in its laboratories. A suitable field former was fabricated and the tests coordinated and financed by ITU were carried out initially by a specialist EMPT company. To date, joints with a helium leak rate of leak in the range 10−8 mBar.litres.s−1 have been achieved. Further refinement of the parameters is now being done in order to improve the weld quality and achieve the target leak-rate of 1.0 × 10−9 mBar.litres.s−1.
An Economic Model of Nuclear Reprocessing Using Vensim
Authors: Brinton Samuel , Tokuhiro Akira
Published: February 2010

Even under the call for solutions to climate change and alternative energy sources to meet increasing energy demands, the imminent “nuclear renaissance” is debated by those who want to know the final destination of spent nuclear fuel. One of the alternatives to direct storage of spent fuel in a geological repository includes partial to full fuel reprocessing such that fission products such as actinides can be removed, as well as the recycling of plutonium and uranium into mixed oxide fuel (MOX). With the anticipated construction of ‘new build’ nuclear power plants (NPPs), as well as the continued operation of the existing fleet, we anticipate that the inventory of spent fuel destined for storage in Yucca Mountain (or similar) will continue to grow. Thus the U.S. DOE is promoting a sensible consideration of reprocessing, burning MOX in existing and near-terms LWRs and continuing R&D on sodium-cooled fast reactors (SFRs) for their eventual commercial introduction. However, countries that have chosen to reprocess are facing high costs and lingering political opposition, while others who have chosen not to reprocess equally face opposition to licensing and operating an adequate federal repository. This research continues ongoing research by the authors on existing and planned realization of NPPs and the associated fuel cycle. That is, we have to date developed models of the construction and decommissioning of NPPs in the U.S., developed an associated model that includes construction of reprocessing facilities, and finally, accounts for the mass flow within the partially closed fuel cycle. From early on, we included the gradual introduction of MOX-burning LWRs and SFRs into the existing and anticipated LWR fleet over the next 100 years. All models were created using Vensim, a software tool that facilitates development, analysis and compartmentalization of dynamic processes with feedback models. Our model has been benchmarked against the MIT and U. Chicago reports on the future of nuclear energy. The current work presents cost estimates and uncertainties assigned to the mass flow model to evaluate the cost of NPP-based electricity generation, with and without a fuel reprocessing. Preliminary results demonstrate that the high cost of reprocessing can be offset by the larger expense of having to construct ‘multiple’ Yucca Mountain-type repositories, under current NPP growth forecasts and insistence of the once-through fuel cycle. Details and results on various, sensible scenarios will be presented.
Study on Boiling Behavior of Solution by Microwave Heating
Authors: Horii Shoichi , Abe Yutaka , Kaneko Akiko , Suzuki Masahiro , Kihara Yoshiyuki , Fujii Kan-ichi
Published: February 2010

At present, several programs for mass production of MOX bulk are on going. The MOX bulk is the capacity of the container to be used to generate MOX fuel. One of them is the selection of the vessel configuration for efficient evaporation of large amount of solution. Two types of vessel as shallow and cylindrical, are under examinations. The shallow vessel has been successfully cleared examinations in moderate running with no flushing, but the amount of products per batch is rather small. On the other hand, the cylindrical vessel is newly proposed with the intent of increasing the production capacity per batch, but has flushing the solution from the top of the vessel by the Microwave Heating (MH). Elucidation on the mechanism of dynamic behavior of the flushing and spilling is our purpose of study. It was carried out by designing a equipment which imitated the practical setup for microwave heating. A visual observation was conducted to clarify the mechanism of dynamic behavior on the sudden nucleation of the solution in the deep cylindrical vessel under MH. The experiment was continued by changing the height and the shape in order to perform the effect evaluation of the size and the shape of vessel.
Spent Resin Hot Compaction at Tihange NPP
Authors: Centner Baudouin , Charpentier David
Published: February 2010

Spent ion exchange media are considered to be problematic waste that, in many cases, requires special approaches and precautions during their immobilization to meet acceptance criteria for disposal. Waste acceptance criteria define, among others, the quality of waste forms for disposal and therefore, will sometimes define appropriate treatment options. Furthermore, the Volume Reduction Factor is a key criterion in Belgium. A technical and economical comparison of industrially available systems, was performed by Tractebel Engineering, and led to decide the installation of a spent resin hot compaction unit at Tihange NPP. Tractebel Engineering is in charge of the project management and takes an active part in the process optimization. The Resin Hot Compaction Process mainly consists of the following steps: wet resin beads filling into a dryer – dewatering – mixing with a binding additive – drying under deep vacuum, until a residual water content of 4% – discharge of the dried waste into compactable drums – drum lid positioning – supercompaction at 1000 tons. In order to avoid cracks on the compacted drum, and surface external contamination from resin leaks, some optimization has been made about additive selection and drum design. Powdered cation exchange resins (as used for BWR condensate polishing) were found to be less compressible than common polymer powders, and to minimize springback as well as crack risks of the drum walls. Placing the compactable drum inside a second, slightly larger drum, also guarantees clean and reproducible pellets with a powder additive proportion limited to 25% of the total waste volume. Without this safety wrapping, the additive proportion must be raised to at least 30%. The obtained matrix inside pellets is a water free, solid but somewhat brittle block. Volume reduction factors between 1.8 and 1.9 are achieved. Special care must be taken about the mortar quality if pellets are to be grouted. This paper describes the Resin Hot Compaction Process application at Tihange NPP.
Time Domain Analysis of the Temperatures in an Electrical Auxiliary Building Room
Authors: Creary Andron , King Matthew F. , Langston Matthew , Kurwitz Cable , Nelson Paul , Yilmaz Fatma
Published: February 2010

A computational fluid dynamics (CFD) model has been developed to predict air temperatures within the switchgear room environment of a nuclear plant’s Electronics Auxiliary Building (EAB). In order to validate the CFD model output, a scale model experiment has been developed using an analytical model to properly scale important EAB room parameters to allow small scale experiments to be performed for validation of the CFD model. The focus of this paper is the development of the methodology used to accurately predict the bulk air temperature or the EAB room. The CFD model is compared to a simple lumped parameter model as well as a scale model experiment. The scaling approach matches the eigenvalues of the lumped parameter model. An experiment based on this scaling approach was performed and compared with CFD output. The time predicted by the CFD model of the Electrical Auxiliary Building room for the average air temperature to increase from 17.7 °C (64 °F) to 40 °C (104 °F) is 23.5 minutes. The lumped parameter analytic solution produces a mean time of 22 minutes. The heat up time for the experiment matches the CFD model providing confidence in the fidelity of the CFD model.
Prony’s Method for BWR Stability Analysis
Authors: Castillo-Dura´n Rogelio , Ortiz-Villafuerte Javier , Rami´rez-Sa´nchez Jose´ Ramo´n , Palacios-Herna´ndez Javier C. , Calleros-Micheland Gabriel , Filio-Lo´pez Carlos
Published: February 2010

Power uprates are becoming a quite viable option both for improved performance and safety of nuclear power plants. Part of the success in power uprates are new designs of fuel assemblies. Nevertheless, power uprates in BWRs has made mandatory new stability analyses since operational and startup maneuvers are not necessarily independent of the performance of such new fuel assembly designs. For example, it has been inferred that certain operational areas, within the before-to-uprate power-flow map, could be reached for the new expected operational conditions, especially during startup, because certain fuel assembly designs have shown faster response to certain neutronic perturbations and their lower coolant flow area. As a consequence, stability monitors are being taken into account as a great help for operation in such special maneuvering conditions. Currently, power oscillations during startup are of not of much concern since boiling boundary length, decay ratio, and other similar parameters are continuously followed by power stability monitors. However, there still exist certain transients, as a recirculation pump trip suddenly occurring at rated operation, in which the instability range can be difficult to determine in real time. One alternative, for a fast analysis of fast events, is Prony’s Method, which is mostly employed in power and energy transport systems problems. This analysis method provides information about stability through a short-time series, in quasi-steady conditions, which may be expected in certain BWR transients. The method computes the complex dampening coefficient, which is the most instable pole in an autoregressive analysis of the time series. However, not many applications in BWR operation have reported and supported to establish the scope of using such analysis for actual events. This work presents the response method to the impulse of an autoregressive model and its relation to the damping coefficient of Prony’s method. It is found a second order behavior with respect to the decay ratio. To support such result, three different noise signals are analyzed where BWR transient events in which low coolant flow and high power conditions are assumed.
Recent Improvements to the Multi-Dimensional Semi-Implicit Two-Phase Flow Solver, CUPID
Authors: Cho H. K. , Park I. K. , Lee H. D. , Jeong J. J.
Published: February 2010

A two-phase analysis code, CUPID, has been developed for a realistic simulation of transient two-phase flows in light water nuclear reactor components. In the CUPID code, a two-fluid three-field model is adopted and the governing equations are solved on unstructured grids for flow analyses in complicated geometries. For the numerical solution scheme, the semi-implicit method of the RELAP5 code, which has been proved to be stable and accurate for most practical applications, was used with some modifications for an application to unstructured non-staggered grids. In the present study, constitutive relations of the two-fluid model were employed including models for an interfacial drag force, interfacial heat transfer, interfacial area, wall heat partition and interfacial non-drag forces. The inter-phase surface topology concept which was proposed for CFD-BWR was applied in order for the present numerical solver to be applicable not only to dispersed flows but also to a flow with a sharp interface. The current paper presents some preliminary calculation results which have been performed for the verification of the inter-phase surface topology. This paper is also concerned with the effects of interpolation schemes on the simulation of two-phase flows. So as to assure a high numerical accuracy, the second-order upwind scheme is implemented into the CUPID code. The calculation results with the second-order upwind scheme were compared with those with the first-order upwind difference. For the comparison, a single-phase laminar flow and a phase separation were simulated. The comparison results among the two interpolation schemes and the exact solutions apparently showed a reduced numerical diffusion with the second order scheme.
Tension Metastable Fluid Detection Systems for Special Nuclear Material Detection and Monitoring
Authors: Lapinskas Joseph R. , Zielinski Stephen M. , Webster Jeffrey A. , Taleyarkhan Rusi P. , McDeavitt Sean M. , Xu Yiban
Published: February 2010

Tension metastable fluid states offer unique potential for radical transformation in radiation detection capabilities. States of tension metastability may be obtained in tailored resonant acoustic systems such as the acoustic tension metastable fluid detector (ATMFD) system or via centrifugal force based systems such as the centrifugal tension metastable fluid detector (CTMFD) system; both under development at Purdue University. Tension metastable fluid detector (TMFD) systems take advantage of the weakened intermolecular bonds of liquids in sub-vacuum states. Nuclear particles incident onto sufficiently tensioned fluids can nucleate critical size vapor bubbles which grow from nanoscales and are then possible to see, hear and record with unprecedented efficiency and capability [1]. Previous work by our group has shown the ability of TMFD systems to detect neutrons with energies spanning eight orders of magnitude with 95%+ intrinsic efficiency [2] while remaining insensitive to gamma photons and also giving directional information [3] on the source of the radiation. In this paper we describe research results with CTMFD systems for use in the detection of key actinide isotopes constituting special nuclear materials (SNMs) in spent fuel. Tests in a CTMFD system demonstrate the ability to detect alpha activity (at ∼100% efficiency) of U-isotopes at concentrations of ∼100 ppb (which is unprecedented and about x100–1000 more sensitive than from conventional liquid scintillation spectroscopy). An inherent capability of TMFD systems concerns on demand tailoring of fluid tension levels allowing for energy discrimination and spectroscopy. This appears especially useful to detect the key isotopes of U and other transuranic isotopes of Pu, Np, Am, and Cm that are at different stages of nuclear fuel reprocessing (i.e. UREX+).
Microwave Based NDE Inspection of HDPE Pipe Welds
Authors: Stakenborghs Robert , Little Jack
Published: February 2010

This paper describes an innovative apparatus and method that has been developed to volumetrically examine dielectric materials, including high density polyethylene piping fusion joints. The method employs a first of a kind apparatus that is based on the creation of an image using electromagnetic energy in the microwave frequency range. The results of comprehensive laboratory testing and actual field inspection of HDPE thermal fusions using the microwave method and apparatus are presented. The specimens examined included both sound thermal fusions and those that included different types of internal flaws that commonly occur in industrial application. Through this research and field application, the apparatus has been shown capable of detecting the presence of internal flaws, such as lack of fusion (i.e. - cold fusion) and inclusions. Also, thickness changes and voids in the HDPE pipe base material were detected and imaged. The results of the NDE technique are compared to mechanical pull test validation. Finally, the scan images of the PE pipe thermal fusions are compared to theoretical fusion microstructure. Relationships between the HDPE fusion images and the basic fusion microstructure are developed.
Transient Response of BWR Core Flow During Simulated Generator Load Rejection Event
Authors: Furuya Masahiro , Hara Takashi , Mizokami Shinya
Published: February 2010

Integral Effects Test (IET) was conducted to investigate the effects of flow redistribution during the generator load rejection event by using the SIRIUS-F facility, which simulates boiling two-phase flow in a BWR core. Owing to the automatic controllers of a recirculation pump inverter and fine-control valves in the facility, the time series of signals of heat flux and mass flux were observed to agree well with those of target rapid flow-decrease events in the previous experimental series. This paper addresses the simulated generator load rejection event, during which the flow and power gradually decrease and the flow takes a turn toward recovery. As a result of the two-parallel channel experiment, mass flux of a hot channel is lower than that of the other during the initial stage. When the void fraction becomes smaller, mass flux of the hot channel is observed to become higher. This phenomenon can be accurately demonstrated with the TRAC-BF1 code as well. The code does, therefore, predict the boiling two-phase flow in a BWR core even at such flow-decrease event. During the event, differential pressure along each channel between the upper and lower plena decreases by several tens of kPa. The relative perturbations of the differential pressure between both channels, however, remain less than 0.4%, which is a significantly small amount. In conclusion, the differential pressures between the upper and lower plena of two-parallel channels are, therefore, identical to each other regardless of the power.
Quantification of the Long-Term Performance of Steels T91 and 316L in Oxygen-Containing Flowing Lead-Bismuth Eutectic at 550°C
Authors: Schroer Carsten , Konys Juergen
Published: February 2010

The long-term performance of ferritic/martensitic steel T91 and austenitic 316L in oxygen-containing flowing lead-bismuth eutectic at 550°C was investigated by exposing the materials for up to 15000 h in the CORRIDA loop. The velocity of the liquid-metal flow was 2 m/s and the concentration of dissolved oxygen averaged 1.6×10−6 mass%. The resulting corrosion processes and products were analyzed and quantified using metallographic methods.
Computational Fluid Dynamics Modeling of PWR Steam Generator Sludge Collector/Loose Parts Weir
Authors: Wang Chunyun , Burr Jason , Karim Naeem
Published: February 2010

The sludge collector is a passive device located in the upper internals of a Pressurized Water Reactor (PWR) recirculating steam generator. The sludge collector’s function is to trap sludge, particulates of small size, i.e., diameters ranging from 1 micron to 64 microns, suspended in the secondary side recirculating flow and entrained from the feedwater inlet flow. Sludge present in the secondary fluid results in degraded thermal performance if deposited on the heat transfer tubing. Another significant contributor to tube degradation is the presence of loose parts in the steam generator. The loose parts weir is a separate passive device located around the sludge collector, which is expected to serve two main purposes: first, to act as a barrier to prevent any loose parts with relative larger size, i.e., greater than 1mm, in the upper internals region from reaching the tube bundle; second, to aid in directing more flow through the sludge collector and improving its efficiency. The loose parts weir is considered for installation in combination with the sludge collector for both new and existing steam generators. Inclusion of the loose parts weir significantly alters the flow field of the sludge collector and thus its performance. A computational fluid dynamics (CFD) study was performed utilizing CFX computer code [1] to explore the performance and interaction of the sludge collector and the loose parts weir. The numerical study results were used to optimize the sludge collector design (inlet and outlet holes sizes, internal configuration) and determine the best placement for the loose parts weir. Particle tracking simulation was also employed to track the path of sludge and loose parts. The sludge deposition on the plates of the collector and the entrapment of the relative larger particles due to the loose parts weir were observed.
Effects of Laser Peening Treatment on High Cycle Fatigue and Crack Propagation Behaviors in Austenitic Stainless Steel
Authors: Ochi Yasuo , Masaki Kiyotaka , Matsumura Takashi , Ikarashi Takaaki , Sano Yuji
Published: February 2010

Laser peening without protective coating (LPwC) treatment is one of surface enhancement techniques using impact wave of high pressure plasma induced by laser pulse irradiation. One of the effects of the LPwC treatment is expected to reduce the tensile residual stress and to induce the compressive residual stress in the surface layer of metallic materials. As a laser has no reaction force due to irradiation and also it has easy characteristics for remote control, the LPwC treatment is practically used as a technique for preventing the stress corrosion cracking (SCC) and for improving the fatigue strength of some structural materials. In this study, high cycle fatigue tests with four-points rotating bending loading were carried out on the non-peened and the LPwC treated low-carbon type austenitic stainless steel 316L in order to investigate the effects of the LPwC treatment on the high cycle fatigue strength and the surface fatigue crack propagation behavior. Two types of specimens were prepared; one was a smooth specimen, the other was a specimen with a pre-crack by the fatigue loading from a small artificial hole. As the results of the LPwC treatment, the high compressive residual stress was induced in the surface layer on the specimens, and the region of the compressive residual stress was about 1mm depth from the surface. The fatigue strength of the LPwC treated SUS316L was remarkably improved during the whole regime of the fatigue life up to the 108 cycles compared with the non-peened materials. Through the fracture mechanics investigation of the pre-cracked materials after the LPwC treatment, it became clear that the fatigue crack propagation was restrained by the LPwC treatment on the pre-cracked region, when the stress intensity factor range ΔK on the crack tip was under the value of 7.6 MPa√m.
Utilization of CMS (Condition Monitoring System) for CBM (Condition-Based Maintenance)
Authors: Morota Hidetsugu , Miyazono Toshimitsu , Okumura Tatsuro
Published: February 2010

CBM (Condition-Based Maintenance) is highly effective to make improvement utilization rate of the electric power facility. In case of CBM, it is often the case that vibration diagnosis for rotating components is direct measurement. CMS (Condition Monitoring System), which is tool for the plant performance analysis and diagnose component fault based on heat and mass balance calculations, is beneficial tool for CBM to proceed more effectively. The work is a suggestion of a new approach for CBM using CMS, provides proof-of-concept demonstration of effectiveness of a foreword looking approach to diagnostic and prognostic technology and shows effectiveness of CMS for CBM including application results.
Simulation on the Thermal Hydraulics of the Coolant in a Pebble Bed Reactor Core by CFD
Authors: Li H. , Qiu S. Z. , Su G. H.
Published: February 2010

The 3-D physical and geometrical model of the water cooled pebble bed reactor (WCPBR) were built using the commercial software ProE and CFX, in which the thermal-hydraulic characteristics of the coolant were studied based on the computational fluid dynamics (CFD) in this work. The velocity field, temperature field and pressure distribution of the coolant are obtained and analyzed in the case of inter-pebble gap and direct contact and its effects on the core safety are also analyzed. In the case of inter-pebble gap, the maximum temperature of the pebble was found in the range of gap, two minima pressures and two maximum velocities were found at the contact points. In the direct contact case, two maximum temperatures were found at the contact points and the maximum velocity was found in the range of gap. From the results, it is evident that the direct contact case is closer to actual case.
Development of Best-Estimate ECCS Evaluation Methodology for APR1400
Authors: Lee Seok-Ho , Kim Mun-Soo , Kim Han-Gon
Published: February 2010

Advanced Power Reactor 1400 (APR1400) is an evolutionary Pressurized Water Reactor (PWR) equipped with such advanced features as the Direct Vessel Injection (DVI), the Fluidic Device (FD) in the Safety Injection Tank (SIT), and the In-containment Refueling Water Storage Tank (IRWST) in the Emergency Core Cooling System (ECCS). To verify the performance of these advanced features, more realistic performance evaluation methodology is desired since existing methodologies use too conservative assumptions which cause negative biases to these features. In this study, therefore, a best estimate evaluation methodology for the APR1400 ECCS under large break loss of cooling accident (LBLOCA) is developed targeting operating license of the Shin Kori 3&4 nuclear power plants (SKN 3&4), the first commercial APR1400 plants. On this purpose, a variety of existing best estimate evaluation methodologies previously used are reviewed. As a result of this review, a methodology named KREM is selected for this study. The KREM is based on RELAP5/MOD3.1K and has been used for Korean operating plants since 2002 when it was first approved by Korean regulation. For this study, RELAP5/MOD3.3 (Patch 3), the latest version of RELAP series is selected since it could appropriately simulate the multi-dimensional phenomena for the APR1400 design characteristics. To quantify the code accuracy, analyses covering experimental data have been performed for 36 kinds of separated effect tests (SETs) and integral effect tests (IETs). The uncertainty in the peak cladding temperature (PCT) of the APR1400 is evaluated preliminarily. Based on the preliminary calculation, final uncertainty quantification and bias evaluation are performed to obtain the licensing PCT for Shin-Kori 3&4 plants and the result shows that the LBLOCA licensing acceptance criteria are well satisfied.
Power Uprate Impact Evaluations on Waste Heat of Nuclear Power Plants in Taiwan
Authors: Peir Jinn-Jer , Shih Chun-Kuan , Pei Bau-Shei , Ferng Yuh-Ming , Hsu Wen-Sheng
Published: February 2010

Power uprate is currently being considered by several nuclear power plant (NPP) utilities as a way to improve plant economy since 1970s. In general there are three types of power uprate: measurement uncertainty recapture (<2%), stretch power uprate (2 to 7%), and extended power uprate (7 to 20%). Prior to the uprate, a thorough evaluation of existing plant equipments and environmental impact must be assessed before a final decision and an optimal selection are made. Upon a power uprate, the power density and thermal release of a nuclear reactor would increase immediately, followed by the impact due to the waste heat from the thermal effluent to the environment. In Taiwan, two BWR NPPs and one PWR NPP are now planning for power uprate. These plants are all located by the side of the seacoast and using marine water as the coolant due to that fact that there is no rivers can supply enough light water near the site of the NPPs in Taiwan. The marine water temperature is high during summer time due to the weather of Taiwan is marine tropical. According to the Effluent Standards of Taiwan’s Environmental Law: for effluents discharged directly into marine waters, the temperature at the discharge point shall not exceed 42 °C; and the temperature difference should not exceed 4 °C for surface water at 500 meters from the discharge point. Therefore, the impact from the waste heat is always a very crucial issue to the power uprate of NPPs in Taiwan. This paper focuses on the environmental impact of power uprate due to waste heat release of NPPs in Taiwan. The evaluations are based on the thermal equilibrium of 100%, 105%, and 110% rated power, respectively. The long term monitor data of marine water temperature are also used to evaluate the impact level from waste heat during normal operation of NPPs. To comply with the Effluent Standards of Taiwan’s Environmental Law, especially in summer, it is needed to solve the thermal effluent’s problem before performing the power uprate of NPP. Adding the dilution pump at each reactor unit is assessed to be a practicable way for resolving such trouble.
Modern Bimodal High Density Polyethylene for the Nuclear Power Plant Piping System
Authors: Haddad Adel N.
Published: February 2010

Originally introduced in the 1990s, bimodal HDPE, pipe resins are still finding new niches today, including even nuclear power plants. HDPE pipe grades are used to make strong, corrosion resistant and durable pipes. High density polyethylene, PE 4710, is the material of choice of the nuclear industry for the Safety Related Service Water System. This grade of polymer is characterized by a Hydrostatic Design Basis (HDB) of 1600 psi at 73 °F and 1000 psi at 140 °F. Additionally bimodal high density PE 4710 grades display >2000 hours slow crack growth resistance, or PENT. HD PE 4710 grades are easy to extrude into large diameter pipes; fabricate into fitting and mitered elbows and install in industrial settings. The scope of this paper is to describe the bimodal technology which produces HDPE pipe grade polymer; the USA practices of post reactor melt blending of natural resin compound with black masterbatch; and the attributes of such compound and its conformance to the nuclear industry’s Safety Related Service Water System.
Control Rod Ejection Accident: Benchmark Solution With FEM Code DiFis
Authors: Zhukov Aleksandr I. , Abdullayev Akif M.
Published: February 2010

A 3D neutron kinetic code DiFis is designed for steady-state and transition calculation for cores with hexagonal fuel assemblies. The code uses traditional time-dependent diffusion equations in a two-group approximation and takes into account six delayed group neutron precursors. A variation finite element method is used for solution of diffusion equations. The thermal-hydraulic model of the code is based on an “average” fuel rod model. Each fuel assembly contains such a rod, which provides heat transfer from a fuel pellet through the gas gap and the cladding to the coolant. The thermal-hydraulic model is used for feedback correction of macroscopic nuclear cross-sections. The code DiFis has been used for the AER-DYN-002 benchmark problem, which describes an eccentric control rod ejection at hot zero power condition in a VVER-440 type reactor. The only feedback mechanism taken into account is a Doppler one. Comparison with well-known nodal kinetic codes shows acceptable accuracy of DiFis. As an example of transition calculation, a VVER-1000 core behavior during fast power reduction is described.
Application of Large-Eddy Simulation to Pressurized Thermal Shock Problem
Authors: Loginov M. S. , Komen E. M. J. , Kuczaj A. K.
Published: February 2010

Life span assessment is a very important issue for the nuclear community. A serious threat to the life span of a Reactor Pressure Vessel (RPV) is an occurrence of the Pressurized Thermal Shock (PTS) during an Emergency Core Coolant (ECC) injection in a loss-of-coolant accident. Traditional system codes fail to predict the complex three-dimensional flow phenomena resulting from such an ECC injection. Improved results have been obtained using Computational Fluid Dynamics (CFD) analysis based on the Reynolds-Averaged Navier-Stokes (RANS) equations. However, it has been shown also that current transient RANS approaches are less capable to predict complex flow features in the downcomer of the RPV. More advanced CFD methods like Large-Eddy Simulation (LES) are required for modeling of these complex flow phenomena in the downcomer. The current paper addresses the feasibility of the application of LES for single-phase PTS. Furthermore, the required grid resolution for such LES analyses is identified by evaluation of solutions on different meshes. A buoyancy-driven PTS experiment has been considered. This experiment has been performed in the 1:5 linear scale Rossendorf Coolant Mixing Model (ROCOM) facility. In the applied numerical model, the incompressible Navier-Stokes equations are solved in the LES formulation, with an additional transport equation for a scalar, which is responsible for driving the embedded buoyancy term in the momentum equations. Instantaneous mixing characteristics are investigated based on evaluation of the scalar concentration. The analysis presented in this paper indicates that the application of LES is feasible nowadays. It is demonstrated that the mixing in the downcomer is quite sensitive to small turbulent disturbances at the ECC inlet, i.e., two simulations performed with slightly different fluctuations at inlet result in substantially different flow in the downcomer. This complicates the analysis of the data from simulations and suggests that validation against experimental data should not be performed using single physical experiment.
Inspection of the Steam Generator Heat Transfer Tubes for FBR Monju Restart
Authors: Takahashi Kenji , Shiina Akira , Onizawa Takahiro , Ibaki Shoji , Yamaguchi Toshihiko , Tagawa Akihiro
Published: February 2010

Japanese prototype FBR Monju will restart its operation for the first time since the occurrence of the sodium leak accident in 1995. The integrity of its components after the long suspension needs to be confirmed before restarting. One of the most important components of Monju is steam generator (SG) unit that was first tested in the 2007.11–2008.3 period during plant-confirmation test. The safety authority required that Monju components should have as same integrity as at the beginning of the first operation in 1994. Now the SG tubes have been installed in the vessel and it is impossible to test them directly. In order to satisfy the safety authority requirement, presumable deteriorations of SG tubes in the long-term suspension were investigated at first and these were limited to only the lack of thickness by corrosion, which are detectable by three tests: eddy current testing (ECT), visual testing (VT), and leak check. These tests were performed with customized procedures and evaluated for appropriate criteria. The ECT was applied to check for the lack of thickness of SG tubes caused by regional corrosions. The ECT probe passes through the full length of a tube with the speed of 200mm/s. Integrity of the tube was checked by amplitude and phase chart. Threshold of the approval thickness was decided based on the tube data at the time of manufacturing. The thickness of the tube is 3.5mm and it cannot be reduced less than 7% lack of the thickness. The characteristic signal at special points like bends or SG support plates were identified by their phase chart. Based on these evaluations, no significant deterioration was detected in SG tubes excepting the special undetectable region. The VT checked for corrosions on the inner surface of tubes. It covers the undetectable regions by ECT. The testing region is between the tube-plate and the bottom U tube along the down comer. The condition of the other regions, subjected to the same environment conditions (temperature, humidity, time not being in service, others) is expected to be similar to the tested region. A customized CCD camera was operated with about 10 mm/s along the tube. The test did not reveal any significant corrosion. The leak test checked for penetrating holes in SG tubes. After keeping a high pressure argon gas outside of SG tube, the nitrogen gas inside of SG was pumped to a mass spectrometer. In this test, no argon gas was detected by the mass spectrometer. After evaluation of all results of three tests, it was confirmed that there is no significant lack of thickness and penetrating holes in SG tubes. Therefore, it was proved that SG heat transfer tubes have the required integrity for reoperation.
Software FMEA Analysis for Safety Software
Authors: Park Gee-Yong , Hur Sup , Kim Dong H. , Lee Dong Y. , Kwon Kee C.
Published: February 2010

This paper describes a software safety analysis for a software code that is installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system. For the ATIP software safety analysis, an overall safety analysis is at first performed over the ATIP software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA, being applied to the ATIP software code which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) which could not be identified during various system tests.
Effectiveness of RCS Depressurization Strategy for Severe Accident Management
Authors: Huh Changwook , Suh Namduk , Park Goon-Cherl
Published: February 2010

In developing the severe accident management guideline (SAMG), it was highly considered to maintain the integrity of reactor pressure vessel (RPV) as a key strategy aimed to reduce the risk of containment failure and fission product releases into the environment effectively. For the operating nuclear power plants with no dedicated safety features for the severe accident management (SAM), the improvement of the current mitigative strategy in SAMG domain could be one of counter-measures to extend the survival time of RPV. In this study, the effectiveness of the RCS depressurization to delay the RPV survival time with different RCS depressurization rate was evaluated for station blackout (SBO) accident assuming only SIT is available for Uljin unit 1 plant by MELCOR 1.8.5 code. According to the analysis results, it was shown that the conditions for RCS depressurization such as depressurization capacity and the time interval are the key elements to extend the RPV integrity in such a way to earn time for restoring the heat sink in order to prevent the accident propagation to the RPV failure.
Framework and Strategies for the Introduction of Best Estimate Models Into the Licensing Process
Authors: Sollima Calogero , Petrangeli Gianni , D’Auria Francesco , Misˇa´k Jozef
Published: February 2010

The current efforts to assure stable, safe and competitive operation of nuclear power plants go together with advances made in accident analysis domain where the deterministic safety analysis is an important instrument for confirming the adequacy and efficiency of provisions for the safety of nuclear power plants. Recently made advances offer two acceptable options for demonstrating that the safety is ensured with sufficient margin: use of best estimate (BE) computer codes either combined with conservative input data or with realistic input data but associated with evaluation of uncertainty of results. The objective, proposed for the present work, is the implementation of the best estimate plus uncertainty (BEPU) method into the licensing process, as developed by the University of Pisa in a PhD thesis. In the thesis activity, more emphasis is given to the study of “input and method” of BE calculations than on “uncertainty evaluation”. In the present paper, a summary of the results achieved is reported.
The Release of Radionuclides in the Laser Decontamination Process
Authors: Herrmann Marion , Lippmann Wolfgang , Hurtado Antonio
Published: February 2010

The decommissioning of nuclear installations requires the decontamination of radioactively contaminated concrete surfaces in order to minimize the amount of radioactive waste to be disposed of as well as the exposure time of the staff during this works. The rapid progress in the development of laser technology has yielded high-performance diode lasers whose radiation can be guided over a long distance by means of glass-fibre optical units. This opens up the possibility of implementing unconventional laser-based decontamination processes. The aim of the method presented here is to combine melting and contactless ablation of a radioactively contaminated concrete surface by means of a laser beam with waste product conditioning. It is intended to design the process in such a way that a maximum of the radioactivity present at the surface is incorporated in the glass melt (= conditioning of waste products). The glassy granulate obtained is very well suited for direct final storage due to its physical and chemical properties. The portion of radioactive isotopes that are released in the process, but not incorporated during the ablation process is selectively deposited in a cooled electro-filter. To prove the effectiveness of the method, research was focused on decontamination experiments conducted on concrete samples contaminated with 137Cs, 60Co and 85Sr. Furthermore, the chemical composition of the concrete samples was varied (quartzitic, quartzitic-calcitic) to take account of the different release conditions in real concrete structures. The experiments showed that 85Sr and 60Co are highly soluble in the glass melt. Their release rate is very low as they have a relatively high boiling point. 137Cs also exhibits a great affinity to the glass melt, but is more easily released again in the high temperature range due to its low boiling point of approx. 700 °C. The released portion of 137Cs is then deposited in the upstream electro-filter. The overall assessment is that the intended decontamination process with simultaneous conditioning of waste products is basically feasible using today’s laser technology. The special advantage can be seen in the great versatility and easy control of the laser unit that is equipped with a fibre-optical system. Furthermore, laser ablation can be set up as a low-dust process, which minimizes problematic secondary contamination.