bluuebunny/crossref_metadata_2025 huggingface dataset:
RELAP5-3D Code Validation for RBMK-1500 Reactors
Authors: Bubelis Evaldas , Kaliatka Algirdas , Uspuras Eugenijus
Published: November 2008

This paper deals with RELAP5-3D code validation through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A successful best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.
Sensitivity Studies on Thermal-Hydraulic Behavior Associated With LOFT Total Loss-of-Feedwater Accident
Authors: Lee Kyu Bok , Lee Heedo , Ko Hee Jin , Seong Ho Je , Jeung Jae Sik , Kim Chang Hyo
Published: November 2008

The purpose of this paper is to study the thermal-hydraulic behaviors associated with the Loss-of-Fluid Test (LOFT) Experiment of total loss-of-feedwater accident and its recovery (Test No. L9-1/L3-3), and to perform sensitivity studies to identify parameters having significant effect on the thermal-hydraulic behaviors predicted by RELAP5/MOD3.3Beta computer code. The Experiment L9-1/L3-3 was a simulation of the total loss-of-feedwater (TLOFW) accident and its recovery modes. Generally, the RELAP5/MOD3.3Beta calculation results were in good agreement with experimental data, even though some differences were also observed. Sensitivity studies showed that several parameters have significant effect on the predicted thermal-hydraulic behaviors. These parameters include the pressurizer spray system loss coefficient, nodalization of the SG secondary side, steam generator (SG) U-tube heat transfer area, heat transfer coefficient from the LOFT main components to the environment, and the power-operated relief valve (PORV) discharge coefficient. The first three parameters are significant during the short-term transient phase, while the last two are significant during the long-term transient phase. Therefore, the five parameters should be carefully modeled to obtain appropriate calculation results for transient types such as the total-loss-of feedwater accident and its recovery modes.
Fuel Element Mixing Promoter Evaluation Using a Thermal Tracing Technique
Authors: Silin Nicola´s , Delmastro Dari´o F. , Juanico´ Luis E.
Published: November 2008

A new thermal tracing technique for the direct measurement of enthalpy mixing between coupled subchannels was used for the performance measurement of different mixing promoters used in nuclear fuel elements (FEs). This new technique allows direct enthalpy mixing measurements through the use of small thermal traces generated by means of a sophisticated superficial heater developed together with a high precision thermometry system. The results show the feasibility of this technique to the direct measurement of interchannel mixing parameter required for numerical codes that are usually employed to study the thermo-hydraulic behavior of new fuel elements designs. In the present work the technique is applied to the quantitative study of the performance of mixing vanes for a compact bundle of rods in completely turbulent flow using water as fluid at ambient conditions. Due to its simplicity this method seems feasible to be a useful fuel element design tool. The experimental results shown asymmetric behaviors not previously reported.
A Technique for Dynamic Corrosion Testing in Supercritical CO2
Authors: Loewen Eric P. , Davis Cliff B. , Shropshire David E. , Weaver Kevan
Published: November 2008

An experimental apparatus for the investigation of the flow-assisted corrosion of potential fuel cladding and structural materials to be used on a fast reactor cooled by supercritical carbon dioxide has been designed. This experimental project is part of a larger research at the Department of Energy being lead by the Idaho National Engineering and Environmental Laboratory (INEEL) to investigate the suitability of supercritical carbon dioxide for cooling a fast reactor designed to produce low-cost electricity as well as for actinide burning. The INEEL once-through corrosion apparatus consists of two syringe pumps, a pre heat furnace, a 1.3 meter long heated corrosion test section, and a gas measuring system. The gas flow rates, heat input, and operating pressure can be adjusted so that a controlled coolant flow rate, temperature, and oxygen potential are created within each of six test sections. The corrosion cell will test tubing that is commercially available in the U. S. and specialty coupons to temperatures up to 600°C and a pressure of 20MPa. The ATHENA computer code was used to estimate the fluid conditions in each of the six test sections during normal operation.
Numerical and Experimental Analysis of Linear Elastic Fracture Mechanics
Authors: Malik Muhammad Afzaal , Sayyed Batool H. , Khushnood Shahab
Published: November 2008

Crack propagation in materials can lead to catastrophic failures. The understanding of crack propagation in materials is of crucial importance in many industries, including nuclear, chemical and aeronautical applications. From an engineering point of view, fracture mechanics is used as a basis for predicting critical crack size, strength of a structure as a function of crack size, inspection requirements pertaining to size of an admissible crack and the period of time between inspections. It is usually required to determine the distribution of stresses and strains in a body that is subjected to external loads or displacements. The closed form solutions for crack propagation are rarely available. A numerical simulation of such problems is therefore required. This research work uses Finite Element Method (FEM) for Linear Elastic Fracture Mechanics (LEFM) to predict Stress Intensity Factor (SIF) for different crack geometries. The validation of the results based on modeling and simulation is carried out through comparison with our experimental investigations for crack propagation using photo-elastic methods.
A Risk-Informed On-Line Test and Maintenance Framework for Redundant Instrumentation and Control Systems in Nuclear Power Plants
Authors: Lu Lixuan , Jiang Jin
Published: November 2008

Deregulation forces Nuclear Power Plants (NPPs) to operate not only safely, but also more economically. Instrumentation and Control (I&C) systems in NPPs play an important role to reduce operational related costs. Therefore, a well-planned test and maintenance (T&M) strategy for I&C systems is more important then ever in this deregulated energy market. This paper presents a general T&M framework for digital I&C systems in NPPs. There are three main parts within this framework: probabilistic safety assessment (PSA), reliability analysis, and costs evaluation. PSA is used to set risk-informed unavailability limits. Reliability analysis is used to derive the relation between the unavailability and the T&M interval. Costs associated with not only T&M, but also reactor mal-shutdowns are evaluated. The Shutdown System Number One (SDSI) in Canadian Deuterium-Uranium (CANDU) based NPPs is used as an example system to illustrate the proposed framework.
Statistical Assessment of Operating Experiences of Power Reactors in France, Japan, and the United States
Authors: Sakamoto Hitoshi
Published: November 2008

Statistical correlations for unavailability of power reactors across countries are sought. France, Japan and the United States, are selected because of their different political climates surrounding nuclear power through the 1990s. Outage data reveal that the dominating type of outage is different in each of the countries in spite of the similar plant types. In France, unplanned, externally caused, partial outages overwhelm other types of outages in number. In Japan, planned outages dominate in terms of number and duration. Unplanned outages are the major type in the U.S. These differences are not only due to technical differences but also to differences in economic and regulatory environments. Results of linear regression analyses suggest that unavailability factors are so random that country of operation, age of reactor and type of reactor cannot predict them well. This finding seems contrasting to an earlier study in the literature.
The Status of the HANARO Class 4 Power Outage
Authors: Kim Hyungkyoo , Jung Hoansung , Wu Jongsup
Published: November 2008

Electric power is essential for all industrial plant. All who use electric power desire a perfect frequency, voltage stability, and reliability all the time. But this cannot be realized in practice because of the many causes of a power supply disturbance that are beyond the control of the utility. Since the first criticality of the HANARO research reactor, the major reasons for reactor trips were system malfunctions and inexperienced operators in the initial stage of its operation. As HANARO is stabilizing, the power supply outage becomes the major reason for a reactor trip. This paper describes the status of power supply outages. This paper deals with not only the outages which have an effect on HANARO operation but also the reasons for the HANARO class 4 power outages. The analysis on the characteristics and the trends of the outages can provide clues to how the outages can be minimized and what the impact of the outages are on the operation.
Core Performance Tests at Low Power for the JOYO Upgrade to MK-III
Authors: Aoyama Takafumi , Schaefer Robert W. , Maeda Shigetaka , Sekine Takashi
Published: November 2008

Many changes were made in the recent upgrade of the experimental fast reactor JOYO to the MK-III design. The core changes which made to achieve a fourfold increase in irradiation capacity are the introduction of a second enrichment zone, an increase in core radius and a smaller decrease in core height. Performance tests done at low power, which focus on the neutronics characteristics, are presented. These tests include nuclear instrumentation system response, the approach to criticality and excess reactivity evaluation, control rod calibrations performed several ways, and isothermal temperature coefficient evaluation. The measurements are emphasized but comparisons with calculated predictions are included. The design predictions are consistent with the performance test results, and all technical safety specifications are satisfied. The JOYO MK-III core will provide enhanced irradiation testing capability, as well as serving as a test bed for improving fast reactor operation, performance and safety.
Modeling the Performance of the BN-350 Cesium Traps
Authors: Wachs D. M. , Romanenko O. G. , Allen K. J. , Zemtsev B. , Maev V. , Dumchev I.
Published: November 2008

A model was generated to evaluate the performance of the cesium trap system used as part of the BN-350 nuclear reactor decommissioning effort in Aktau, Kazakhstan. A series of cesium traps were designed to remove cesium from the reactor primary circuit sodium and piping to minimize personnel radiological exposure while draining and disposing of the sodium during subsequent decommissioning activities. Cesium was trapped by extracting sodium from the primary circuit, passing it through a block of reticulated vitreous carbon (RVC), and returning the cleaned sodium to the primary circuit. Cesium has a strong affinity for adsorbing onto the surface of carbon and RVC in particular has been proven to be an effective cesium collecting material. A model was created to predict and help explain the performance of the system during operation. The model is effectively split into two components, the reactor primary system component and the cesium trap RVC component. The temperature dependent parameters governing the behavior of the system were continuously updated. The model output effectively described the behavior observed during the operation of the first three cesium traps (the final four traps were used to clean a variety of ancillary system whose initially conditions were not well known and, consequently, did not lend themselves to being modeled). The defining system operating conditions were the primary sodium temperature of approximately 290 °C, trap RVC temperatures range of 150 °C to 260 °C, and dissolved cesium concentrations in the primary sodium ranging from 7.2×10−3 Ci/kg to 2.4×10−4 Ci/kg.
Effect of Inlet Flow Resistance on Instabilities During Startup Transient in Natural Circulation BWRs
Authors: Kuran Selim , Sun Xiaodong , Xu Yiban , Yoon Ho Jun , Cheng Ling , Revankar Shripad T. , Ishii Mamoru , Wang Weidong
Published: November 2008

The startup transient was simulated in the Purdue University Multidimensional Integral Test Assembly (PUMA) facility for natural circulation boiling water reactors (BWRs) with different values of core inlet flow loss coefficient. The simulation was performed by means of pressure and power scaling, considering the core inlet subcooling and the structure-stored energy differences between the model (PUMA) and the prototype (Simplified BWR). A novel core inlet plate was designed and installed so that the loss coefficient could be easily changed without disassembling the reactor pressure vessel. The simulation experiments demonstrated the pattern of the geysering oscillations during the single-to-two-phase natural circulation transition phase. The core inlet flow loss coefficient was not observed to have a significant effect on the geysering oscillations.
Seismic Re-Evaluation of the Tarapur Atomic Power Plants 1 and 2
Authors: Ingole S. M. , Santosh Kumar B. , Gupta Sushil , Singh U. P. , Giridhar K. , Bhawsar S. D. , Samota A. , Chhatre A. G. , Dixit K. B. , Bhardwaj S. A.
Published: November 2008

Two Boiling Water Reactors (BWR) of 210 MWe each at Tarapur Atomic Power Station, Units-1&2 (TAPS-1&2) were commissioned in the year 1969. The safety related civil structures at TAPS had been designed for a seismic coefficient of 0.2g and other structures for 0.1g. The work of seismic re-evaluation of the TAPS-1&2 has been taken up in the year 2002. As two new Pressurized Heavy Water Reactor (PHWR) plants of 540 MWe each, Tarapur Atomic Power Project Units-3&4 (TAPP-3&4), are coming up in the vicinity of TAPS-1&2, detailed geological and seismological studies of the area around TAPS-1&2 are available. The same free-field ground motion as generated for TAPP-3&4 has been used for TAPS-1&2. The seismic re-evaluation of the plant has been performed as per the procedure given in IAEA, Safety Reports Series entitled “Seismic Evaluation of Existing Nuclear Power Plants”, and meeting the various codes & standards, viz., ASME, ASCE, IEEE standards etc. The Safety Systems (SS) and Safety Support Systems (SSS) are qualified by adopting detailed analysis and testing methods. The equipment in the SS and SSS have been qualified by conducting a walkdown as per the procedure given in Generic Implementation Procedure, Dept. of Energy (GIP–DOE), USA. The safety systems include the systems required for safe shutdown of the plant, one chain of decay heat removal and containment of activity. The safety support systems viz., Electrical, Instrumentation & Control and systems other than SS & SSS have been qualified by limited analysis, testing and mostly by following the procedure of walkdown. The paper brings out the details of the work accomplished during seismic re-evaluation of the two units of BWR at Tarapur.
Experimental Study on Post-Dryout Heat Transfer of Dispersed Flow in Vertical Narrow Annuli
Authors: Myint Aye , Tian Wenxi , Li Zhihui , Qiu Suizheng , Jia Dounan , Su Guanghui
Published: November 2008

Forced convective post dryout heat transfer in narrow channel with 1.2 mm gap has been experimentally investigated with deionized water. The experiment was carried out with pressure ranging from 1.38 to 5.9 MPa and low mass velocity from 52.9 to 84.2 kg/m2 s. The experimental data were compared with well known empirical correlations such as Groeneveld, Polimik, Miropolskiy and Slaughterbeck and it was found that these correlations could not predict very well in narrow annular gap at low mass velocity. Based on the experimental data, the heat transfer coefficient increases with increasing heat flux, mass flux and pressure. A new empirical correlation for narrow annuli at low mass velocity was then developed which has a good agreement with the experimental data.
Extensions to SCDAP/RELAP5/ATHENA for Analysis of HTGRs
Authors: Harvego Edwin A. , Siefken Larry J.
Published: November 2008

The SCDAP/RELAP5/ATHENA code was extended to enable the code to perform transient analyses of two leading candidates for the next generation of nuclear power plants, namely the Pebble Bed and Block-Type (prismatic) High Temperature Gas Reactors (PB-HTGRs and BT-HTGRs). Models in five areas of reactor behavior were implemented into the code. First, models were implemented to calculate the transfer of decay heat by conduction and radiation from the center of the heterogeneous cores of these reactors to their outer surfaces. Second, models were implemented to calculate the transfer of decay heat from the outer surfaces of these reactor cores to ultimate heat sinks such as the atmosphere. Third, models were implemented to calculate the flow losses and convective heat transfer in the core of a PB-HTGR. Fourth, models were implemented to calculate the oxidation of graphite due to the ingress of air. Fifth, a model was implemented to calculate the ingress of air into the reactor core by molecular diffusion from the point of break in the coolant system. The extended code was applied to the analysis of conduction cooldown accident in a BT-HTGR. The analysis indicated that the new models in SCDAP/RELAP5/ATHENA are properly interfaced with the previously developed and assessed fluid behavior models in the code. Preliminary results indicate that post-shutdown decay heat can be adequately removed from BT-HTGRs by natural circulation of air from the atmosphere.
Fundamental Study on Two-Phase Flow Characteristics in Chimney of Pb-Bi Cooled Direct Contact Boiling Water Fast Reactor
Authors: Akashi Toyou , Yamada Akira , Takahashi Minoru
Published: November 2008

The lead-bismuth-steam two-phase flow in the chimney of the lead-bismuth cooled direct contact boiling water fast reactor (PBWFR) was simulated experimentally using water-air two-phase flow system and analytically using the multi-dimensional two-fluid model computer program ACE-3D for the design of the ring header nozzles and chimney of the PBWFR. It has been found that the downward flow appeared in the chimney according to the analytical results. The analytical simulation will be verified by the experiment data, and the air injection flow rate, and the number, caliber and arrangement of injection nozzles in the ring header will be chosen adequately to improve the performance of the lift pump. Also, the analytical program will be extended for the simulation of a steam-Pb-Bi flow system.
Fundamental Research on the Cooling Characteristic of a Passive Containment Cooling System
Authors: Kawakubo Masahiro , Inaba Noriaki , Kikura Hiroshige , Aritomi Masanori , Yamauchi Toyoaki
Published: November 2008

The objective of this experimental study is to clarify the heat transfer characteristics of the Passive Containment Cooling System (PCCS) with vertical heat transfer tubes for investigating the influence of non-condensable gas on condensation. Furthermore, hence we obtained new experimental correlation formula to calculate the transients in system temperature and pressure using the simulation program of the PCCS. The research was carried out using a forced circulation experimental loop, which simulates atmosphere inside PCCS with vertical heat transfer tubes if a loss of coolant accident (LOCA) occurs. The experimental facility consists of cooling water supply systems, an orifice flowmeter, and a tank equipped with the heat transfer pipe inside. Cooling water at a constant temperature is injected to the test part of heat transfer pipe vertically installed in the tank by forced circulation. At that time, the temperature of the cooling water between inlet and outlet of the pipe was measured to calculate the overall heat transfer coefficient between the cooling water and atmosphere in the tank. Thus, the heat transfer coefficient (hc) between heat transfer surface and the atmosphere in the tank considering the influence of the non-condensable gas was clarified. An important finding of this study is that the amount of condensation in the steamy atmosphere including non-condensable gas depends on the cooling water Reynolds number, especially the concentration of non-condensable gas that has great influence on the amount of condensation.
A Direction to Adopting Risk-Informed and Performance-Based Regulation (RIPBR) in Korea
Authors: Kim Woong Sik , Sung Key Yong , Lee Chang Ju , Choi Young Hwan , Choi Yong Seog , Kim Hho Jung
Published: November 2008

The USNRC has pursued the incorporation of new regulatory approach of risk-informed and performance-based regulation (RIPBR) into nuclear safety regulation, as an alternative to improve existing nuclear safety regulation of nuclear power plants, which is deterministic and prescriptive. It focuses on the use of risk insight from probabilistic safety assessment (PSA). Recently, it becomes necessary to find a way to improve regulatory efficiency and effectiveness in order to cover the increasing regulatory needs in Korea. Also, the utility has optimized design and operation of the plant using PSA insight and equipment performance information. According to the increase of the necessity for regulatory improvement using risk and performance information, KINS (Korea Institute of Nuclear Safety) is developing, as a part of mid- and long-term project of Nuclear R&D program, how to adopt the RIPBR in Korean nuclear regulation. This paper presents the interim result of the study that comprises the assessment of feasibility to adopt the RIPBR, and basic directions and principles for implementing RIPBR model. It is concluded that adopting RIPBR is essential for risk management of the plant, public acceptance to nuclear safety, improvement of existing regulation, minimization of unnecessary regulatory burden, and effective use of regulatory resources, etc. Three basic directions and several principles that are necessary to implement RIPBR model are identified from the study. The application of the directions and principles to the assessment of RIPBR model to be established in the near future is expected to result in making the adoption of new regulatory system more objective and consistent.
A Neutron Amplifier: Prospects for Reactor-Based Waste Transmutation
Authors: Blanovsky Anatoly
Published: November 2008

A design concept and characteristics for an epithermal breeder controlled by variable feedback and external neutron source intensity are presented. By replacing the control rods with neutron sources, we could maintain good power distribution and perform radioactive waste burning in high flux subcritical reactors (HFSR) that have primary system size, power density and cost comparable to a pressurized water reactor (PWR). Another approach for actinide transmutation is a molten salt subcritical reactor proposed by Russian scientists. To increase neutron source intensity the HFSR is divided into two zones: a booster and a blanket with solid and liquid fuels. A neutron gate (absorber and moderator) imposed between two zones permits fast neutrons from the booster to flow to the blanket. Neutrons moving in the reverse direction are moderated and absorbed in the absorber zone. In the HFSR, neptunium-plutonium fuel is circulated in the booster and blanket, and americium-curium in the absorber zone and outer reflector. Use of a liquid actinide fuel permits transport of the delayed-neutron emitters from the blanket to the booster, where they can provide additional neutrons (source-dominated mode) or all the necessary excitation without an external neutron source (self-amplifying mode). With a blanket neutron multiplication gain of 20 and a booster gain of 50, an external neutron source rate of at least 1015 n/s (0.7MW D-T or 2.5MW electron beam power) is needed to control the HFSR that produces 300MWt. Most of the power could be generated in the blanket that burns about 100 kg of actinides a year. The analysis takes into consideration a wide range of HFSR design aspects including the wave model of observed relativistic phenomena, plant seismic diagnostics, fission electric cells (FEC) with a multistage collector (anode) and layered cathode.
Nuclear Instrumentation Systems in Prototype Fast Breeder Reactor
Authors: Vijayakumaran P. M. , Nagaraj C. P. , Paramasivan Pillai C. , Ramakrishnan R. , Sivaramakrishna M.
Published: November 2008

The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection & Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident.
Generation of Consistent Data for Neutron Kinetics Models of the GRS System Code ATHLET
Authors: Langenbuch S. , Velkov K.
Published: November 2008

The paper describes the first experience at GRS with a switch algorithm built into the system code ATHLET, which allows to turn to point kinetics or 3D calculations with the neutronics core model QUABOX/CUBBOX and vica versa. The heart of the algorithm is the neutronics data generation code SIGMAS, developed and validated at GRS. Its basic characteristics and possibilities of applications are briefly described. As a demonstration of the algorithm, the results of two boron transient calculations performed with the switch coupling are presented and discussed.
Use of Percentile Rank Sum Method in Identifying Repetitive High Occupational Radiation Dose Jobs in a Nuclear Power Plant
Authors: Cho Young Ho , Ko Hyun-Seok , Kim Suk-Hoon , Kang Chang Sun , Moon Joo Hyun , Kim Kyoung Doek
Published: November 2008

The cost-effective reduction of occupational radiation dose (ORD) at a nuclear power plant could not be achieved without going through an extensive analysis of accumulated ORD data of existing plants. Through the data analysis, it is required to identify what are the jobs of repetitive high ORD at the nuclear power plant. In this study, Percentile Rank Sum Method (PRSM) is proposed to identify repetitive high ORD jobs, which is based on non-parametric statistical theory. As a case study, the method is applied to ORD data of maintenance and repair jobs at Kori units 3 and 4 that are pressurized water reactors with 950 MWe capacity and have been operated since 1986 and 1987, respectively in Korea. The results was verified and validated, and PRSM has been demonstrated to be an efficient method of analyzing the data.
Enhanced Defense in Depth: A Fundamental Approach for Innovative Nuclear Systems Recommended by INPRO
Authors: Kuczera Bernhard , Juhn Poong-Eil
Published: November 2008

In May 2001, the IAEA initiated the “International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)”. Having in mind that nuclear power will be an important option for meeting future electricity needs, the scope of INPRO covers nuclear reactors expected to come into service in the next fifty years, together with their associated fuel cycles. INPRO has been implemented in two phases. In the framework of its Phase 1A, the task group Safety of Innovative Nuclear Systems examined the needs to be met by innovative nuclear systems and defined, in a hierarchical order, a corresponding set of Basic Principles, User Requirements, and Criteria. In this context, an enhanced defense-in-depth strategy is considered as a dominant part in the fundamental safety approach. For each of the five levels of protection, a user requirement indicates the direction of the enhancement (e.g.: no need for evacuation, even in case of severe accidents) and, in general, more independence of the levels from each other shall be achieved.
Lessons Learned From GEN I Carbon Dioxide Cooled Reactors
Authors: Shropshire David E.
Published: November 2008

This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.
Realistic Safety Margin Analysis of “MONJU” Based on Plant Performance Measurements
Authors: Yamada Fumiaki , Kitamura Kenji
Published: November 2008

Safety of the Prototype Fast Breeder Reactor ‘MONJU’ was confirmed at the Installation Permission stage by a series of plant behavior analyses which covered anticipated accidents hypothesizing conservative conditions for various safety reviews. In order to quantify the conservativeness of these analytical conditions, the same series of accidents have been simulated based on actual plant performance data measured during the commissioning tests, to predict more realistic plant accident behaviors. The new analytical results show that the safety margin of the maximum mid-wall temperature of fuel-pin clad increases about 100°C compared to the formerly derived value from the safety review. The main reason of this difference is that the safety review assumes the reactivity worth of the control rods in case of plant trips is at the minimum level required for a reactor shutdown. This recent revised margin has been evaluated to be equivalent to about 10% of the potential increase of plant thermal output.
Review of Practice for Deeply Embedded/Buried NPP Structures Subject to Seismic Loadings
Authors: Xu J. , Miller C. , Hofmayer C. , Graves H.
Published: November 2008

Motivated by many design considerations, several conceptual designs for advanced reactors have proposed that the entire reactor building and a significant portion of the steam generator building will be either partially or completely embedded below grade. For the analysis of seismic events, the soil-structure interaction (SSI) effect and passive earth pressure for these types of deeply embedded structures will have a significant influence on the predicted seismic response. Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to assess the significance of these proposed design features for advanced reactors, and to evaluate the existing analytical methods to determine their applicability and adequacy in capturing the seismic behavior of the proposed designs. This paper summarizes a literature review performed by BNL to determine the state of knowledge and practice for seismic analyses of deeply embedded and/or buried (DEB) nuclear containment type structures. Included in the paper is BNL’s review of the open literature of existing standards, tests, and practices that have been used in the design and analysis of DEB structures. The paper also provides BNL’s evaluation of available codes and guidelines with respect to seismic design practice of DEB structures. Based on BNL’s review, a discussion is provided to highlight the applicability of the existing technologies for seismic analyses of DEB structures and to identify gaps that may exist in knowledge and potential issues that may require better understanding and further research.
CFD Simulation of Steam Generator Tube Rupture Thermal-Hydraulics
Authors: Maslovaric Blazenka , Stevanovic Vladimir , Prica Sanja , Stosic Zoran
Published: November 2008

The tube rupture accident is one among the most risk-dominant events at the nuclear power plants. Several steam generator tube rupture accidents have occurred at the plants in the past. In this paper the Computational Multi-Fluid Dynamics (CMFD) investigation of the horizontal steam generator thermal-hydraulics during the tube rupture accident is performed. A guillotine of a steam generator U-tube is assumed with choked flow from the primary to the secondary side of the steam generator. Predicted are water and steam velocity fields, steam volume fraction distribution on the steam generator secondary (shell) side, as well as the swell level increase. Obtained multidimensional results are a support to the safety analyses of the steam generator tube rupture accident. Also, they serve as benchmark tests for an assessment of the applicability of one-dimensional horizontal steam generator models, developed by standard safety codes. Numerical simulation is performed with the multidimensional multi-fluid modelling approach. The two-phase flow around steam generator tubes in the bundle is modelled by the porous media approach. Interfacial mass, momentum and energy transfer is modelled with the closure laws, where some of them are specially developed for the conditions of the two-phase flow across tube bundles. The governing equations are solved with the SIMPLE type pressure-correction method that is derived for the conditions of multi-phase flow conditions.
Safety of Nuclear Reactors: Part A — Unsteady State Temperature History Mathematical Model
Authors: El-Shayeb Mohamed , Yusoff Mohd. Zamri , Boosroh Mohd Hariffin , Bondok Ali , Ideris Fazril , Abu Hassan Saiful Hasmady
Published: November 2008

A nuclear reactor structure under abnormal operations of near meltdown will be exposed to a tremendous amount of heat flux in addition to the stress field applied under normal operation. Temperature encountered in such case is assumed to be beyond 1000°C. A mathematical model has been developed for the fire resistance calculation of a concrete-filled square steel column with respect to its temperature history. Effects due to nuclear radiation and mechanical vibrations will be explored in a later future model. The temperature rise in each element can be derived from its heat balance by applying the parabolic unsteady state, partial differential equation and numerical solution into the steel region. Calculation of the temperature of the elementary regions needs to satisfy the symmetry conditions and the relevant material properties. The developed mathematical model is capable to predict the temperature history in the column and on the surface with respect to time.
Effect of Wall Surface Wettability Due to Radiation Induced Surface Activation on Quenching Process
Authors: Tamura Naohisa , Hazuku Tatsuya , Takamasa Tomoji , Mishima Kaichiro , Mitsutake Toru , Morooka Sinichi
Published: November 2008

This report presents an experimental result on quenching which affected by the Radiation Induced Surface Activation. Increasing of quenching velocity requires that the cooling liquid can contact the heated surface, or a high wettability of the surface when quenching initiates. From this point of view, an experimental study to investigate the relationship between surface wettability and quenching was performed by use of an oxide semiconductor-coated material after γ-ray irradiation. The results showed that highly surface wettability or highly hydrophilic condition of a simulated fuel rod made of SUS was achieved, and the quenching velocities were increased up to 15% after 300 kGy 60Co γ-ray irradiation.
New ASME Rules for Spent Nuclear Fuel and High Level Radioactive Material and Waste Storage Containments
Authors: Hill Ralph S. , Foster Gerald M.
Published: November 2008

In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) will be published. The new Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. Some time after publication, CC N-717 will be incorporated into the body of the Code. This paper provides an informative insight to the Code Case so that Owners, regulators, designers, and fabricators have a more comprehensive understanding.
Integration of Proton Beam Collimation Scheme in the Spallation Neutron Source (SNS) Accumulator Ring Including Dose and Activation Estimates
Authors: Simos Nicholas , Ludewig Hans , Raparia D. , Catalan-Lasheras N. , Cousineau S. , Davino D.
Published: November 2008

This paper details the integration scheme as well as the induced activation by the proton beam clean-up system (collimation) in the accumulator ring section of the Spallation Neutron Source (SNS) accelerator complex. Specifically, the results of the optimization study in terms of satisfying both the optics of the proton beam and the minimization of activation of the accelerator components as well as of the surrounding structure have guided both the design of the components and their integration and are presented in this paper. The resulted collimation scheme is a two-stage clean-up system consisting of proton beam halo intercepting scrapers and appropriate fixed aperture absorbers. The accumulator ring structure consists of the High Energy Beam Transfer (HEBT) line which receives the 1 GeV proton beam from the SNS LINAC accelerator, the accumulator ring itself which compiles the micro-pulses into the final 60 Hz pulse, and the RTBT line that transfers the final proton pulse to the accelerator target. Collimation takes place in all three ring components and along respective straight sections with the exception of the off-momentum particle clean-up in the HEBT line in which off-momentum protons are guided to a stationary absorber outside the transfer line. Given that the activation issues of the collimating structures themselves as well as of the nearby accelerator components (mainly magnets) are similar in all sections of the ring, the activation of components in the ring clean-up system will be discussed in detail in the following sections.
On Synergy Between a High Temperature Gas-Cooled Reactor and a LNG Vaporization Plant
Authors: Morrow Charles W.
Published: November 2008

The Liquefied Natural Gas (LNG) chain of processes consumes the equivalent of 10% of initial natural gas flow for liquefaction, transportation and regasification of the natural gas. It is possible with the right process to recover some of this lost investment during the regasification process. The High Temperature Gas Cooled Reactor (HTGR) nuclear power plant appears to possess the characteristics needed to accomplish this recovery. This synergy of processes and fluid properties between an LNG regasification plant and an HTGR provides an opportunity to enhance an already efficient nuclear power generation scheme. Boiling LNG (112 K) provides an ideal cold side heat sink for the helium based Brayton cycle of the HTGR. Helium remains in the gas phase at these low temperatures. The resulting large temperature difference (1000 K) between the high temperature and low temperature sides of a thermal cycle means Carnot efficiencies approach 90%. Achievable efficiencies approach 77%, an increase from 48% for current ambient temperature cooled HTGR designs. Thus a LNG/HTGR plant can deliver half again more power for similar capital investments and operating costs. In addition, boiling LNG with helium saves fuel gas costs for the regasification plant. This paper will show that this combination is feasible and economic. Since both processes are designed to run at maximum capacity, duty cycles and plant availability criteria match. For coastal locations, both processes possess similar site selection criteria. Finally, combining the processes will impose no unmanageable safety constraints on either system and in fact could make safe operation easier to attain. This paper will provide general overviews of an HTGR power plant and of the LNG-to-market sequence, concentrating on regasification plants. The paper will then describe a process that combines an HTGR power plant with an LNG regasification facility to the advantage of both. At full load, the economic benefit for a dual installation supporting what would be a 1.1 GWe power plant before improvement would be approximately $423 million per year.
Seismic Fragility Analysis of a Water Storage Structure
Authors: Bhargava K. , Ghosh A. K. , Ramanujam S.
Published: November 2008

The present paper is concerned with the seismic response and fragility evaluation of a water storage structure. Seismic analysis has been carried out considering the hydrodynamic effects of the contained water. The various parameters that could affect the seismic structural response include material strength of concrete, structural damping available within the structure and the normalized ground motion response spectral shape. Based on this limited case study; the seismic fragility of the structure is developed as families of conditional probability curves plotted against peak ground acceleration (PGA) at the location of interest. The procedure adopted incorporates the various randomness and uncertainty associated with the parameters under consideration.
Subcooled Boiling Data From Rod Bundles
Authors: Karoutas Zeses E. , Sung Yixing , Chang Yutung R. , Kogan Gennady A. , Joffre Paul F.
Published: November 2008

This paper provides single and two phase rod bundle data to support verification of heat transfer models being used in steaming rate and crud model predictions for rod bundles. The effort to summarize this work was supported by the EPRI Robust Fuel Program and is defined in more detail in EPRI report 1003383. Subcooled boiling tests were performed by Combustion Engineering (CE) in early 1980s to provide insight on heavy crud deposits and fuel failures observed on peripheral rods for bundles in Maine Yankee cycle 4. Two 5×5 tests were performed at the Columbia University Heat Transfer Research Facility simulating the peripheral region of adjacent CE 14×14 fuel bundles for two different perimeter strip geometries. The test conditions were at typical reactor pressure, temperature, and heat flux. The rods were 7’ in heated length and were electrically heated with a uniform axial power shape. There were no mixing vanes on the spacer grids. Thermocouples were placed on the hot rod in the center of the test section and on an adjacent rod at 4 different axial levels. Thermocouples were also located in the center of the subchannels at the end of the test section. Boiling curves were generated over a range of test conditions (system pressure, inlet temperature, and flow rate) by plotting rod surface temperature versus heat flux. The boiling curves covered single phase, subcooled boiling, bulk boiling and DNB conditions. The data from the boiling curves were reduced and evaluated with the VIPRE thermal hydraulic code. Clad temperature predictions were made with VIPRE based on available heat transfer correlations for comparison to clad temperature measurements. These heat transfer correlations include the Dittus Boelter correlation for single phase flow, the Jens Lottes, Thom and Chen correlations for two phase flow conditions (subcooled boiling). The VIPRE predictions of the hot rod average surface temperature, based on the Dittus-Boelter correlation with a grid enhancement factor for single-phase forced convection and the Thom correlation for nucleate boiling, gave the best agreement with the rod bundle test data among all the available modeling options. It was concluded that current heat transfer models used in TH codes, are adequate for average steaming rate calculations supporting Axial Offset Anomalies (AOA) evaluations, as long as the appropriate grid enhancement factor is utilized for the spacer grids in the analysis. However, further testing and modeling may be needed to simulate local grid effects and hot spots downstream of spacer grids.
Scoping Studies for a Small Modular Lead-Cooled Fast Reactor
Authors: Yu Chenggang , Smith Michael A. , Feldman Earl E. , Yang Won Sik , Sienicki James J.
Published: November 2008

A scoping design study has been carried out of the feasibility of a small, 25 MWt (∼10 MWe), modular lead-cooled fast reactor coupled to an advanced power converter consisting of a gas turbine Brayton cycle that utilizes supercritical carbon dioxide as the working fluid. Major constraints of the study are an ultralong 20 year core lifetime, near zero reactivity burnup swing over the core lifetime, Pb primary coolant natural circulation heat transport, road transportability of plant modular assemblies including the reactor and guard vessels, and high Brayton cycle power conversion efficiency. It is found that the goal of a near zero reactivity burnup swing implies a low core power density that results in an unacceptably low discharge burnup.
Modelling and Dynamics Analysis of Heat Exchanger as a Distributed Parameter Process
Authors: Savic B. , Debeljkovic D. Lj.
Published: November 2008

On the basis accepted and critically clarified assumptions, a non–linear and afterwards linearized mathematical model of fuel oil cooling chamber has been developed in engineering sense sufficiently correct. The model is in the form of set of partial differential equations with constant coefficients. Using the appropriate numerical simulation of the results derived, the dynamic of this process has been shown in the form of appropriate transient processes responses which quite well correspond to the real process behavior.