Commercial Nuclear Power Plants Research Articles

Assessment of RELAP5 model for the University of Massachusetts Lowell Research Reactor

RELAP5 is a system code developed at the Idaho National Environmental and Engineering Laboratory for thermal hydraulic analysis of nuclear reactors. The code RELAP5 is widely used for safety analysis studies of commercial nuclear power plants. However, recent released version of RELAP5/3.2 and over present significant capabilities for analysis of nuclear reactor research systems. As a contribution to the assessment of RELAP5/3.3 for research reactor safety analysis, experimental data from the University of Massachusetts Lowell Research Reactor - UMLRR are used. The UMLRR is a 1 MW light water moderated and cooled, graphite-reflected, open-pool type research reactor. This paper presents the development and the validation of a UMLRR-RELAP model using experimental data. For this purpose, a series of experiments were performed for benchmarking RELAP5 calculations for research reactor systems. As a result of this study, the UMLRR nodalization is shown to be representative of the experimental data reactor behavior.

First Decommissioning Project of Commercial Nuclear Power Plant in Japan

First Decommissioning Project of Commercial Nuclear Power Plant in Japan

THE NUCLEAR HOT POTATO

OVER THE PAST FEW WEEKS, Energy companies and Congress have advanced plans to build a centralized interim storage facility for spent fuel from commercial nuclear power plants. Such projects are gai...

Predicting nuclear power-plant operator performance using discrete event simulation

A study was conducted to evaluate the use of discrete event simulation (DES) to predict human performance in a nuclear power plant control room environment. Computer simulation models of two disturbance scenarios were built using a simulation software program, Micro Saint. In parallel, data were also collected at a full-scope training simulator at the Halden man-machine laboratory (HAMMLAB) in Halden, Norway, using crews of commercial nuclear power plant operators from the Loviisa nuclear power plant in Loviisa, Finland. Comparisons were made between predicted operator performance data generated by the simulation models and crew performance in the HAMMLAB experiment to determine the degree of agreement between the simulated data and the data from operators. The models were then used to extrapolate advanced control room conditions and alarm systems that were not tested in the HAMMLAB experiment. This report summarizes these findings and provides recommendations for improvements to the DES approach for use by a regulatory agency.

Emergency management: Consequence management decision making

Emergency managers have a dilemma in deciding what to do when there is an emergency that affects their community. Those who have habitual hazards in their community are basically prepared. When a tornado is sighted in “tornado alley,” everyone knows what to do. When a hurricane is coming to shore along the Florida and Texas Gulf Coast, there are basic emergency steps to follow. But in this time of new and more challenging risks, we need a better system to coordinate community emergency decision making, no matter what the hazard. A simple solution is to adopt the four-level emergency event classification system that is already in use with communities with commercial nuclear power plants.

Simple cost risk-benefit calculation: nuclear plant backfit analysis

The radiological risks include the probabilities and consequences of severe accidents at commercial nuclear power plants. The purpose of this paper is to present a valuation on the expected person-rem or societal committed dose from major accidents to identify whether back-fits in design modification can be supported. The approach of project evaluation by engineers of amortization is somewhat modified for these probabilistic events. A sample calculation is presented along with examples from published analytical calculations for several US light water reactors.

Assessing risk: the role of probabilistic risk assessment (PRA) in patient safety improvement.

Morbidity and mortality due to "medical errors" compel better understanding of health care as a system. Probabilistic risk assessment (PRA) has been used to assess the designs of high hazard, low risk systems such as commercial nuclear power plants and chemical manufacturing plants and is now being studied for its potential in the improvement of patient safety. PRA examines events that contribute to adverse outcomes through the use of event tree analysis and determines the likelihood of event occurrence through fault tree analysis. It complements tools already in use in patient safety such as failure modes and effects analyses (FMEAs) and root cause analyses (RCAs). PRA improves on RCA by taking account of the more complex causal interrelationships that are typical in health care. It also enables the analyst to examine potential solution effectiveness by direct graphical representations. However, PRA simplifies real world complexity by forcing binary conditions on events, and it lacks adequate probability data (although recent developments help to overcome these limitations). Its reliance on expert assessment calls for deep domain knowledge which has to come from research performed at the "sharp end" of acute care.

Uncertainty Analysis of Memory Based Sensor Validation Techniques

This paper presents an analytical derivation and analysis of the uncertainty of the Multivariate State Estimation Technique (MSET). Like all other nonparametric techniques, MSET uncertainty consists of two parts: bias and variance. Bias is a systematic error in MSET inference and practically not computable and non-removable, but when properly regularized it is usually very small with respect to the variance when properly regularized. Variance, on the other hand, represents variability of the MSET estimate due to random noise in the data and can be estimated in real time. All the derivations and results are obtained for the inferential case. The MSET cost function is also derived which shows that MSET minimizes a weighted least squares cost function with weighting affected by the MSET memory matrix. The parallels between MSET and more traditional kernel techniques, namely kernel regression, are drawn and it is shown that MSET is a special type of kernel regression algorithm. The final section presents the results of the MSET uncertainty analysis for real world data obtained from a commercial nuclear power plant.

Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor

Significant amounts of nuclear wastes consisting of plutonium, minor actinides and long lived fission products are produced during the operation of commercial nuclear power plants. Therefore, the destruction of these wastes is very important with respect to public health, environment and also the future of nuclear energy. In this study, transmutation of minor actinides (MAs) discharged from LMFBR spent fuel in a high power density fusion reactor has been investigated under a neutron wall load of 10 MW/m2 for an operation period of 10 years. Also, the effect of MA percentage on the transmutation has been examined. The fuel zone, containing MAs as spheres cladded with W-5Re, has been located behind the first wall to utilize the high neutron flux for transmutation effectively. Helium at 40 atm has been used as an energy carrier. At the end of the operation period, the total burning and transmutation are greater than the total buildups in all investigated cases, and very high burnups (420–470 GWd/tHM) are reached, depending on the MA content. The total transmutation rate values are 906 and 979 kg/GWth year at startup and decrease to 140 and 178 kg/GWth year at the end of the operation for fuel with 10% and 20% MA, respectively. Over an operation period of 10 years, the effective half lives decrease from 2.38, 2.21 and 3.08 years to 1.95, 1.80 and 2.59 years for 237Np, 241Am and 243Am, respectively. Total atomic densities decrease exponentially during the operation period. The reductions in the total atomic densities with respect to the initial ones are 79%, 81%, 82%, 83%, 85% and 86% for 10%, 12%, 14%, 16%, 18% and 20% MAs, respectively.

ステンレス鋼

Many kinds of stainless steels were developed for the usage in each industrial field. The research and development on stainless steel in these 30 years were described. A new refining process of stainless steel made it possible to reduce an impurity element to extra low contents, and to develop new stainless steels, e. g., many kinds of ferritic stainless steels. Furthermore, many of these stainless steels were standardized to the Japanese Industrial Standard.Super stainless steel with high corrosion resistance was developed as a roof material at a seaside area. New ferritic stainless steels with high thermal fatigue resistance were adopted to an automobile as an exhaust manifold. The suitable alloys from 13Cr stainless steel to high-alloyed steel were developed for tubings of oil and gas wells where the corrosion resistance largely depends on the contents of CO2 and/or H2S. Many corrosion problems occurred in a commercial nuclear power plant. The new corrosion phenomena were studied and elucidated by many researchers.

Recent trends in repair and refurbishing of steam turbine components

The repair and refurbishing of steam generator components is discussed from the perspective of repair welding philosophy including applicable codes and regulations. Some case histories of repair welding of steam generator components are discussed with special emphasis on details of repair welding of cracked steam turbine blades and shrouds in some of the commercial nuclear power plants using procedures developed.

Tritium on Ice: The Dangerous New Alliance of Nuclear Weapons and Nuclear Power

In December 1998, Energy Secretary Bill Richardson announced that the U.S. planned to begin producing tritium for its nuclear weapons in commercial nuclear power plants. This decision overturned a fifty-year policy of keeping civilian and military nuclear production processes separate. Tritium, a radioactive form of hydrogen, is needed to turn A-bombs into H-bombs, and the commercial nuclear power plants that are to be modified to produce tritium are called ice condensers. This book provides an insider's perspective on how Richardson's decision came about, and why it is dangerous.Kenneth Bergeron shows that the new policy is unwise not only because it undermines the U.S. commitment to curb nuclear weapons proliferation but also because it will exacerbate serious safety problems at these commercial power facilities, which are operated by the Tennessee Valley Authority and are among the most marginal in the United States. The Nuclear Regulatory Commission's review of the TVA's request to modify its plants for the new nuclear weapons mission should attract significant attention and opposition.Tritium on Ice is part expose, part history, part science for the lay reader, and part political science. Bergeron's discussion of how the issues of nuclear weapons proliferation and nuclear reactor safety have become intertwined illuminates larger issues about how the federal government does or does not manage technology in the interests of its citizens and calls into question the integrity of government-funded safety assessments in a deregulated economy.

Historical perspective on failure rates for US commercial reactor components

To support the development of probabilistic risk assessments of US commercial nuclear power plants, significant effort has been expended to develop generic failure rates for components. Generic failure rates indicate industry-average performance of components, rather than component performance at a specific plant. Most publicly available, generic failure rate databases are typically based on data collected in the 1970s and 1980s for US nuclear power plants. Recent data analysis programs sponsored by the US Nuclear Regulatory Commission and data collection programs sponsored by the Institute of Nuclear Power Operations provide an opportunity to compare more recent failure rate estimates with those obtained in the 1970s and 1980s. These recent results indicate that many component generic failure rates are now lower than observed in the 1970s and 1980s. Suggestions for up-to-date failure rates are presented. Also, failure to run rates for standby components are presented for both short- and longer-term run times.

ICONE11-36622 PRELIMINARY DESIGN OF THE POWER CONVERSION UNIT WITH DIRECT GAS-TURBINE CYCLE FOR THE HTR-10

After the first phase of the 10MW high temperature gas-cooled test module reactor (HTR-10) with steam-turbine cycle, the preliminary design of the power conversion unit (PCU) with direct gas-turbine cycle for the next phase of HTR-10 project was finished by the Institute of Nuclear Energy Technology (INET) of Tsinghua University and OKBM (Russia). The main characteristics of the preliminary design, including the PCU structure, the thermal circuit and equipment parameters are described detailed in this paper. Implementation of research program for gas-turbine cycle on the HTR-10 reactor will make it possible to gain necessary experience in gas-turbine cycle design for high-temperature gas-cooled reactor and creation of equipment of large commercial nuclear power plant in the future.

Silica transport in the high-level radioactive waste repository, Yucca Mountain, Nevada

Abstract The United States is currently planning to bury high-level radioactive wastes from commercial nuclear power plants and from weapons production in a mined repository in tuff deposits under Yucca Mountain, Nevada. Radioactive decay in these wastes will create steep thermal gradients in the rocks surrounding the repository. This heat might cause the pore water in these tuffs to evaporate, migrate as vapor to cooler areas, and condense to form a solution that can aggressively dissolve the surrounding tuff. Under certain circumstances the silica-rich, bicarbonate solutions produced by this process could migrate under the influence of gravity back toward the canisters where the water would again evaporate leaving amorphous silica behind to form a low permeability silica cap above the repository. Over time, a perched water table containing carbonate and colloidal silica-rich brine could develop above the repository. Thermoelastic fracturing of the silica cap might allow this brine to flow into the repository. We developed a vertical thermal gradient experiment that simulates the silica dissolution and transport processes and used it to determine that each Joule of heat can leach and transport as much as 1.85(+ or -0.56)X10 (super -8) g of silica from a crushed sample of the Topopah Springs Member of the Paintbrush Tuff. Based on this value and the thermal models of Moujaes and Lei (1995), we estimate that the heat from each canister of waste might transport as much as 31.7 kg of silica during the first 100 years and as much as 136 kg of silica during the first 1,000 years after emplacement. Assuming a reasonable fracture density and a 30 mX30 m area of influence for each canister, we estimate that the silica cap above the repository could grow to be 2.1 mm thick after 100 yr and 9.1 mm thick after 1,000 yr.

Reactivity, Isotopic, and Thermal Steady-State Analysis of Homogeneous Thoria-Urania Fuel

Recent investigations into the performance and economics of mixed thoria-urania fuel cycles demonstrate potential advantages at high burnup. Initial neutronic and fuel behavior calculations for several ThO2-UO2 mixtures being considered for use in commercial nuclear power plants are described.The Monte Carlo N-Particle-Origen2 Coupled Utility Program (MOCUP) was used to analyze the reactivity characteristics and isotopic concentrations of unit fuel pins/cells and lattice/assembly models as a function of burnup and reactivity. Neutronic results for a three-batch 6-yr cycle for each of three proposed ThO2-UO2 mixtures with the UO2 enriched to 19.5% 235U are presented. Neutronic results show that fuels fabricated from ThO2-UO2 mixtures can reach an average discharge burnup of up to 70 MWd/kgHM, which will increase the time between refueling and decrease the production of weapons-grade plutonium by a factor of 3 as compared to all-urania fuel.A version of FRAPCON-3, modified to handle pure thoria and ThO2-UO2 mixtures, was used for the fuel performance and behavior calculations. The new version called FRAPCON-3Th includes the updated material property models for thermal conductivity, specific heat capacity, emissivity, thermal expansion, modulus of elasticity, and melting temperature to predict fuel behavior for pure ThO2 or ThO2/UO2 mixed fuel. For a concentration of 75% ThO2/25% UO2, initial fuel performance parameters (peak centerline temperature, gap conductance, thermal expansion, etc.) predicted operating conditions are better than those of current UO2 fuel. A ThO2-ThO2/UO2 thermal conductivity model is still in the development stage. For all fuel calculations, an interim model that interpolates between the Belle and Berman predicted thermal conductivity using a correction factor for radiant heat transport and the MATPRO-predicted thermal conductivity for UO2 was applied.

Neutron measurements at nuclear power reactors [55

Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D 2O-moderated 252Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participated in neutron assessments at a number of LWR facilities to characterize neutron radiation fields and to evaluate the responses of plant dosimeters and survey instruments. In these studies, the tissue equivalent proportional counter was used for measuring neutron dose and dose equivalent rates, and multisphere spectrometers were used to measure energy distributions. The use of these instruments in LWR work locations is usually difficult because of extreme environmental conditions such as high temperatures. These studies have confirmed the presence of low-energy neutron fields in most work locations. The studies have also found that albedo dosimeters used at power plants typically overrespond significantly when using a calibration based on californium exposures. Survey instruments also respond highly in typical LWR environments.

YUCCA MOUNTAIN REPOSITORY FUROR

A DRAFT FEDERAL REPORT has found significant technical and managerial problems in the Department of Energy program to construct a high-level radioactive waste repository at Nevada's Yucca Mountain. The draft General Accounting Office (GAO) report was leaked to the media as Energy Secretary Spencer Abraham was poised to announce whether the project, which is intended to begin taking spent fuel from commercial nuclear powerplants by 2010, will proceed. A favorable recommendation to move ahead was expected early next year from Abraham. In light of the report, however, the Nevada congressional delegation is urging President George W. Bush to delay a Yucca Mountain decision until technical questions can be resolved. For his part, Abraham last week blasted the views expressed in the draft report as well as its release. According to the draft, DOE lacks technical information needed to make a recommendation to proceed as well as data necessary to apply for a Nuclear Regulatory ...

The Design of Human–Machine Interface for Accident Support in Nuclear Power Plants

As feedback from Three Mile Island No. 2, a large amount of human–machine interface (HMI) design has been proposed to support operators during accidents by presenting information on plant status, some of which is implemented in commercial nuclear power plants (NPPs). However, it has not yet been discussed what role HMI should play under conditions where operators must take action as instructed under emergency operating procedure. Regarding this principal issue, an HMI design is proposed together with specific screen images. The advantage of this design is based on the involvement of an ex-operator in NPP in this paper. The participation of users in the design of HMI has been rare. Accordingly the proposed HMI is comprehensive for average operators and is expected to be acceptable for future implementation in commercial NPPs. The effectiveness of the proposed HMI has been examined in a small experiment.

技術報告 ３４２． 原子炉施設の廃止措置計画策定および管理のための計算システムの開発 東海発電所の解体作業に関するプロジェクト管理データの試算

Data on the Japan Power Demonstration Reactor dismantling activities were analyzed to be applied to planning and evaluation of decommissioning reactor facilities. These were then combined with the computer systems for planning and management of reactor decommissioning (COSMARD) as a database to estimate decommissioning project management data such as manpower needs and radiation exposure to workers. The Tokai Power Station decommissioning program, which will be conducted by the Japan Atomic Power Company as the first project on decommissioning a commercial nuclear power plant in Japan, was characterized by evaluating the project management data using COSMARD and the database. It was found from the study that COSMARD and unit productivity factors were applicable to planning and evaluation of reactor decommissioning programs.