This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The ...

Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

Information is presented concerning the scenario for core degradation; early core degradation in small break loss-of-coolant accidents; and containing a degraded core accident. (ERA citation 08:026288)

Sandia National Laboratories is participating in several NRC-sponsored programs to study severe accident phenomenology. Part of that effort involves the combined use of the computer codes MARCH and HECTR to examine hydrogen behavior during severe accident...

Engineering calculations are performed to predict the temperatures of the gas and structures along the primary-coolant system during postulated severe reactor accidents. The calculations cover the time span between the beginning of core uncovery and core ...

Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core.

The containment response to a postulated core meltdown accident in a PWR Ice Condenser Containment and a BWR Mark III Containment was examined to see if the WASH-1400 containment failure mode judgment for the Surry large, dry containment and the Peach Bot...

If a reactor accident leads to core melt, interaction of this material with coolant can produce hydrogen by steam oxidation of the metallic content of the melt. Experimental results are presented for hydrogen generation from both explosive and non-explosi...

Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the ...

Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

Information is presented concerning comments on reactor safety by governmental agencies and civilian organizations; reactor safety study methodology; consequence model; probability of accident sequences; and various accident conditions.

Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; ...

This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the ...

The Severe Accident Sequence Analysis (SASA) Program, sponsored by the US Nuclear Regulatory Commission (NRC), is addressing a number of accident scenarios that potentially pose a health hazard to the public. Two of the scenarios being analyzed in detail ...

A wide range of accident scenarios for an ice-condenser containment have been analyzed using the combined capabilities of MARCH and HECTR. The results are found to be sensitive to the combustion modeling parameters. No conclusive statements can be drawn with respect to partial pre-inerting. It appears that the possible benefits afforded by a reasonably sized vent are much less ...

The purpose of this work is to describe some of the research being performed at ORNL in support of the effort to describe, as realistically as possible, fission product source terms for nuclear reactor accidents. In order to make this presentation manageable, only those studies directly concerned with fission product behavior, as opposed to thermal hydraulics, ...

This paper highlights the interim results of ASEP. ASEP is one of many programs sponsored by the Nuclear Regulatory Commission (NRC) to study various aspects of severe core damage accidents. The results of ASEP will ultimately focus on identifying the most likely accident sequences and the high risk accident sequences for most light ...

An understanding of the freezing of molten debris on cold core structures following a hypothetical core meltdown accident in a light water reactor (LWR) is of importance to reactor safety analysis. The purpose of the present investigation was to analyze the transient freezing of the molten debris produced in a severe reactivity initiated accident (RIA) ...

Boiling water reactor (BWR) fuel at 56 to 61 GWd/tonne U was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated accident conditions. Current Japanese 8 x 8 type Step II BWR fuel from Fukushima Daini Unit 2 was refabricated to short segments, and thermal ...

The methodology developed in support of the US Nuclear Regulatory Commission's (NRC's) evaluation of severe accident risks in NUREG-1150 is noted. This paper discusses the results. The principal technical analyses for NUREG-1150 were performed at Sandia National Labs. under the Severe Accident Risk Reduction Program and the ...

Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by...

Laboratory experiments on the thermal interaction of simulated light water reactor (LWR) fuel melts and water are summarized. Their purpose was to investigate the possibility of steam explosions occurring for a range of hypothetical accident conditions. Pressure, temperature, hot liquid motion and cold liquid motion were monitored during the experiments.

The U.S. Nuclear Regulatory Commission sponsors a broad range of research on the response of nuclear fuel assemblies to normal, off-normal, and accident conditions in light-water reactors. The paper reviews the current status of three Zircaloy cladding research programs in progress at the Oak Ridge National Laboratory and presents some preliminary results from each.

Collability models for post-accident nuclear reactor debris are compared with data for volume-heated debris under various conditions (uniform debris on an impermeable support for various fluids and pressures, debris with forced liquid flow from below, and stratified debris). Comparisons are presented in graphical form and by calculating the average error in fitting available ...

This report summarizes the results of the Bulletins and Orders Task Force review of IE bulletins, Commission Orders, and the NRR generic evaluation of feedwater transients, small-break loss-of-coolant accidents, and other TMI-2 related events in operating plants to confirm or establish the bases for their continued safe operation. The results of this evaluation are presented.

Information is presented concerning chemical forms of fission product iodine in the primary circuit; chemical forms of fission product iodine in the containment building; summary of iodine chemistry in light water reactor accidents; and impact of the radiodine source term on the potassium iodide issue.

This report presents an assessment of the adaptability of EPRI's one- and two-dimensional STEALTH computer codes to perform transient fuel rod analysis. The ability of the STEALTH code to simulate transient mechanical or thermomechanical loss-of-coolant accident is described. Analytic models of one- and two-dimensional formulations and features included in the ...

In the unlikely event that molten core debris escapes the reactor pressure vessel, the interactions of the debris with concrete and structural materials become the driving forces for severe accident phenomena. The Ex-vessel Core Debris Interactions Program at Sandia Laboratories is a research effort to characterize the nature of these interactions and the magnitude of ...

The physico-chemical I exp 131 species were determined in the stack discharge of a BWR power plant, which had been shut down because of an accident, and in the stack discharge of a PWR power plant. The duration of measurement was 8 and 49 weeks respective...

The guide delineates acceptable design limits and appropriate combinations of loadings associated with normal operation, postulated accidents, and specified seismic events for the design of Class 1 linear-type component supports as defined in Subsection NF of Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The guide applies ...

A computer program modelling the interaction between molten core materials and structural concrete is being developed to provide a capability for making quantitative estimates of reactor fuel-melt accidents. The principal phenomenological models, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. A code test comparison calculation ...

The objective of the Core Thermal Model Development program currently in progress at the Battelle Pacific Northwest Laboratory is to develop numerical simulation methods for evaluating thermal hydraulic performance of water-cooled nuclear reactor cores and other components during postulated accident conditions. The primary thrust of this program to date has been continued ...

The exact failure pressures of containment designs are unknown. Probabilistic risk assessment (PRA) analyses of risk from nuclear power plants use estimates of the failure presusre to calculate the relative probabilities of release of radioactive material by various containment failure scenarios (or modes). The containment failure mode determines the consequences of the ...

In the last few years, credit for fuel burnup has been allowed in the design and criticality safety analysis of high-density spent-fuel storage racks. Design and operating philosophies, however, differ significantly between pressurized water reactor (PWR)- and boiling water reactor (BWR)-type plants because: (1) PWR storage pools ...

This paper describes preliminary results from which finite element calculation results are used in conjunction with analytical calculation results to predict failure in different LWR vessel designs during a severe accident. Detailed analyses are being performed to investigate the relative likelihood of a BWR vessel and drain line penetration to fail during ...

Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the...

The containment response to a postulated core meltdown accident in a PWR Ice Condenser Containment and a BWR Mark III Containment was examined to see if the WASH-1400 containment failure mode judgment for the Surry large, dry containment and the Peach Bottom Mark I containment are likely to be appropriate for ice-condenser and Mark III ...

Sandia National Laboratories is participating in several NRC-sponsored programs to study severe accident phenomenology. Part of that effort involves the combined use of the computer codes MARCH and HECTR to examine hydrogen behavior during severe accidents. These codes have been applied to an ice-condenser containment. Results indicate the importance of ...

Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy ...

In the analysis of degraded core accidents, the two major sources of pressure loading on light water reactor containments are: steam generation from core debris-water thermal interactions; and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work ...

A general analysis approach which includes the dynamic time-history analysis and the static analysis is presented. The secondary system piping is that piping which excludes the primary coolant loop in the Pressurized Water Reactor (PWR) plant, and the main steam, feedwater and reactor recirculation piping in the Boiling Water Reactor (BWR) plant. Piping ...

The FRAP-T5 transient fuel rod behavior code, with its automated uncertainty analysis subcode, was used to estimate the uncertainty of best estimate calculations simulating five selected reactor accident and transient events. Code input was supplied by the four US reactor vendors. Response surface methodology was used to estimate calculation uncertainty. Fifteen fuel rod ...

Many of the high-consequence accident sequences in the Reactor Safety Study (WASH-1400) involved containment failure by overpressurization. One way to relieve pressure and avoid containment failure is by venting of the containment atmosphere. The study undertook to develop some quantitative insight into the potential effect on public risk of venting and filtering the ...

Flooding of the cavity of a Pressurized Water Reactor (PWR) and the drywell of a Boiling Water Reactor (BWR) is one of the many accident management strategies being proposed to manage severe accidents in light water reactors. In this work the effect of external cooling on the thermal behavior of the vessel lower ...

Because the containment structure is the last barrier against the release of radioactivity, an assessment was undertaken to identify the design weaknesses and estimate the margins of safety for the SEP containments under the postulated, combined loading conditions of a safe shutdown earthquake (SSE) and a design basis accident (DBA). The design basis ...

The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters c...

Tests data are presented for BWR and PWR rods in test HI-4 and test HI-5. Operating conditions fission product release data are included.

... Title : Innovative Robotics and Ultrasonic Technology at the Examination of Reactor Pressure Vessels in BWR and PWR Nuclear Power Stations,. ...

Underwater lighting equipment and techniques for BWR and PWR fuel handling operations are described. (DCC)

... Title : Effect of Sensitization and Cold Work on Stress Corrosion Susceptibility of Austenitic Stainless Steels in BWR and PWR Conditions,. ...

Vaporized structural materials form the bulk of aerosol particles that can transport fission products in severe LWR accidents. As part of the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, a model has been developed based on a mass transport coefficient to describe the transport of materials from the surface of a molten ...

The knowledge of the leakage behavior of containments beyond design conditions is required for the evaluation of severe accident mitigation strategies, risk studies, emergency preparedness planning, and siting. Since each containment building has a large number of penetrations, these represent a large number of potential leak paths from the containments. Four NRC programs - ...

The leakage behavior of containments beyond design conditions and knowledge of failure modes is required for evaluation of mitigation strategies for severe accidents, risk studies, emergency preparedness planning, and siting. These studies are directed towards assessing the risk and consequences of severe accidents. An accident ...

To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the ...

Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be ...

This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, ...

Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe ...

A modular computational system known as the Water Reactor Analysis Package (WRAP) is being developed at the Savannah River Laboratory for analysis of loss of coolant accidents (LOCA's) and other transients in water reactor systems. At this time, WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a ...

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed ...

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed ...

An analysis of spent fuel heatup following a hypothetical accident involving drainage of the storage pool is presented. Computations based upon a new computer code called SFUEL have been performed to assess the effect of decay time, fuel element design, storage rack design, packing density, room ventilation, drainage level, and other variables on the heatup characteristics of ...

Many postulated nuclear reactor accidents result in high-temperature dryout or film boiling within the nuclear core. In order to mitigate potential fuel rod damage or rod failure, safe or lower fuel rod temperatures must be reestablished by promoting coolant/cladding contact. This process is commonly referred to as quenching or rewetting, and often, these terms are not ...

One serious concern during the accident at Three Mile Island resulted from the generation of a large amount of hydrogen gas and its migration to the containment building. The fear was that combustion of this hydrogen could produce pressures that would breach containment. The Center for Nonlinear Studies at Los Alamos is sponsoring an effort to assess the Laboratory's ...

Recent work aimed at correctly describing nonequilibrium vapor generation rates in flashing liquids in decompressing flows similar to those which might be encountered in a loss of coolant accident in a nuclear reactor is summarized. Analysis is reviewed which describes the flashing inception superheat in terms of the turbulence intensity for a given expansion rate and initial ...

Predicted are the Sequoyah and McGuire ice-condenser-containment atmosphere pressure ad temperature responses during degraded-core accidents that involve the release and combustion of large quantities of hydrogen. Predictions are based on (1) an appropriately modified version of the US Nuclear Regulatory Commission (NRC) subcompartment analysis code COMPARE and (2) a ...

Significant retention of fission products in the primary system can occur during severe reactor accidents. Some of these retention processes have been identified, including their reaction rates and reaction products. The reactions investigated include: CsI, CsOH, and tellurium with the structural materials Inconel 600 and 304 stainless steel, tellurium with tin, silver and ...

Sandia National Laboratories is presently involved in several NRC-sponsored experimental projects to provide data that will help quantify the threat of hydrogen combustion during LWR accidents. One project, which employs several experimental facilities: is the Variable Geometry Experimental System (VGES). The purpose of this paper is to present the experimental results from ...

The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant-accident (LOCA) situations in light-water reactors (LWR's). The effects of internal pressure, heating rate, axial constraint, and ...

The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as ...

Sandia National Laboratories is currently involved in a number of experimental projects to provide data that will help quantify the threat of hydrogen combustion during nuclear plant accidents. Several experimental facilities are part of the Variable Geometry Experimental System (VGES). The purpose of this report is to document the experimental results from the first round of ...

As part of Sandia's Severe Accident Risk Reduction (SARR) Program for the U.S. Nuclear Regulatory Commission, the cost-benefit tradeoffs of filtered-vented containment (FVC) systems, hydrogen control systems, alternate decay heat removal systems, and a variety of other reactor modifications designed to reduce the risk from severe accidents have ...

The goal of this document is the organization of operating teams and procedures in accident and emergency situations in EDF PWR units. (Atomindex citation 24:067289)

The main phenomena leading to hydrogen generation in the containment of a PWR in accident conditions are described in this report. They are the following: water radiolysis, zirconium cladding oxidation and oxidation of the aluminium of internal structures...

The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, was examined in O2 enriched BWR conditions (8 ppm O2) and in typical PWR conditions. Cracking susceptibility in BWR ...

Studies suggest that the risk of severe accidents during low power operation and/or shutdown conditions could be a significant fraction of the risk at full power operation. The Nuclear Regulatory Commission has begun two risk studies to evaluate the progression of severe accidents during these conditions: one for the Surry plant, a pressurized water ...

Studies suggest that the risk of severe accidents during low power operation and/or shutdown conditions could be a significant fraction of the risk at full power operation. The Nuclear Regulatory Commission has begun two risk studies to evaluate the progression of severe accidents during these conditions: one for the Surry plant, a pressurized water ...

One consequence of the accident at the Three Mile Island Unit 2 (TMI-2) nuclear power plant is a realization by the nuclear power technical community that there is a need for calculational tools that can be used to analyze the TMI-2 accident and to investigate hypothetical situations involving degraded light-water reactor (LWR) cores. As a result, there ...

A methodology for determining the probability spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated ...

A methodology for determining the probability spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated ...

The Thermal Fuels Behavior Program (TFBP) of EG and G Idaho conducts fuel behavior research in the Power Burst Facility (PBF) at INEL and at the Halden Reactor in Norway. The fuels behavior research in the PBF is directed toward providing a detailed understanding of the response of light water reactor (LWR) nuclear fuel assemblies to off-normal and hypothesized accident ...

In the unlikely event of a core meltdown accident, an important safety issue is the potential for steam explosions and their effects on the accident progression. Steam explosion phenomena can be divided into three stages: (a) mixing of the molten fuel and water; (b) triggering and spatial propagation of rapid fuel fragmentation through the fuel-coolant ...

The potential benefits of filtered-vented containment systems as a means for mitigating the effects of severe accidents are analyzed. Studies so far have focused upon two operating reactor plants in the United States, a large-containment pressurized water reactor and a Mark I containment boiling water reactor. Design options that could be retrofitted to these plants are ...

Pressurized ejection of melt from a reactor pressure vessel has been identified as an important element of a severe reactor accident. Copious aerosol production is observed when thermitically generated melts pressurized with nitrogen or carbon dioxide to 1.3 to 17 MPa are ejected into an air atmosphere. Aerosol particle size distributions measured in the tests have modes of ...

The risks associated with the transport of spent nuclear fuel by truck and rail have been reexamined and compared to results published in NUREG-O170 and the Modal Study. The full reexamination considered transport of PWR and BWR spent fuel by truck and rail in four generic Type B spent fuel casks. Because they are typical, this paper presents results only ...

This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident m...

This report documents an analysis of the radiological source terms for severe accidents in Light Water Reactors (LWR) using the Industry Degraded Core Rulemaking Program (IDCOR) integrated accident analysis methodology. The analytical study is an on-going effort sponsored by EPRI to investigate the sensitivity of the Modular Accident ...

The influence of chemistry on in-vessel severe accident phenomena in integral effects in-pile tests is reviewed. In-vessel severe accident chemistry involves high temperature interactions between the materials used in the tests; namely, fuel rods, control materials, spacer grids, and steam. The influence of chemistry on the release and transport of fission ...

Studies and operating experience suggest that the risk of severe accidents during low power operation and/or shutdown (LP/S) conditions could be a significant fraction of the risk at full power operation. Two studies have begun at the Nuclear Regulatory Commission (NRC) to evaluate the severe accident progression from a risk perspective during these ...

This paper describes preliminary results from which finite element calculation results are used in conjunction with analytical calculation results to predict failure in different LWR vessel designs during a severe accident. Detailed analyses are being performed to investigate the relative likelihood of a BWR vessel and drain line penetration to fail during ...

The MELCOR 1.8.4 code Bottom Head package has been applied to simulate two reactor cavity flooding scenarios for when the corium material relocates to the lower-plenum region in postulated severe accidents. The applications were preceded by a review of two main physical models, which highly impacted the results. A model comparison to available bibliography models was done, ...

The decontamination tests have been carried out on samples coming from representative specimens from primary circuit of the PWR and on samples coming from the emergency feed water piping of the German BWR (Isar). The oxide found in PWR primary loops can o...

During the past four years, the ORNL BWRSAT Program has developed a series of increasingly sophisticated BWR secondary containment models. These models have been applied in a variety of studies to evaluate the severe accident mitigation capability of BWR ...

Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined ...

The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they ...

This paper presents an assessment of the time-dependent degradation or aging of selected major light water reactor components and structures. The stressors, possible degradation sites and mechanisms are identified for several major LWR components: pressurized water reactor (PWR) reactor pressure vessels, PWR containment structures, PWR ...

As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the ...

This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core ...

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed ...

The concrete structure represents a line of defense in safety assessment of containment integrity and possible minimization of radiological releases following a reactor accident. The penetration study of hot UO/sub 2/ particles into limestone concrete and basalt concrete highlighted some major differences between the two concretes. These included penetration rate, melting and ...

A temperature measurement error analysis was performed for the Type S (0.25-mm-diam, bare-wire) and Type K (0.71-mm-diam, sheathed) thermocouple circuits used to measure the temperature of the Zircaloy-clad, electrically heated fuel-rod simulators in the Multirod Burst Test program (MRBT) at Oak Ridge National Laboratory (ORNL). An important objective of the MRBT program is to improve the ...

Some postulated accident scenarios lead to a picture with most of the fuel and some structural steel sealed inside the original core volume by frozen plugs at the original core entrance and exit. eventually the fuel will either melt out or blow out of such a sealed volume but in the interim between plug formation and eventual dispersion the question remains How will the fuel ...

The proposed load combination project has the following overall objectives: develop a methodology for appropriate combination of dynamic loads for nuclear power plants under normal plant operation, transients, accidents, and natural hazards; establish design criteria, load factors, and component service levels for appropriate combinations of dynamic loads or responses to be ...

THERM models a reactor fuel rod during the reflood phase of the loss-of-coolant-accident (LOCA) analysis. The purpose of its development is twofold: to formulate a physically founded reflood heat transfer correlation and to identify important parameters governing the reflood process, a process that now limits the operation of many nuclear plants. THERM performs a ...

This article reports that a major effort by the utility industry has helped reverse the upward trend in radiation exposure of its personnel. Between 1983 and 1985, the median exposure decreased from 1030 to 680 rem at boiling-water reactor (BWR) plants, and from 500 to 395 rem at pressurized-water reactor (PWR) plants. Current exposures are at the lowest ...

The Super SARA Test Program (SSTP) is a major European Economic Community effort to study light-water reactor safety during large- and small-break loss-of-coolant accident (LOCA) events in the ESSOR reactor. The SSTP will simulate small-break LOCA's by producing slow temperature ramps to high fuel rod temperatures (approx. 2000/sup 0/C). These experiments will provide ...

Major activities included: (1) the vital area analyses of operating reactor facilities (both PWR and BWR); (2) assistance to NRC for evaluation of the Three Mile Island accident; (3) further development and testing of the ADPATH (adversary paths) subroutine for finding single-target theft and multiple-target sabotage paths in a ...

During a hypothetical core meltdown accident, an important safety issue to be addressed is the potential for steam explosions. This paper presents analysis and modelling of experimental results. There are four observations that can be drawn from the analysis: (1) vapor explosions are suppressed by noncondensible gases generated by fuel oxidation, by high ambient pressure, and ...

A theoretical model has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO/sub 2/-base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel ...

Light water reactor degraded core accident sequences have been postulated which lead to deposition of superheated debris beds in the reactor cavity. The debris beds are assumed to be cooled by an overlying pool of cooling water. The quenching of the debris is predicted to pressurize the containment building as a result of the steam generated during the quench process. ...

The High Temperature Fission Product Program, an experimental and analytical effort to produce thermodynamic and reaction-rate data on fission product species during severe LWR accidents, is reviewed. The paper discusses: (1) a facility where non-radioactive isotopes of fission products react with steam, reactor materials, and hydrogen for periods of time from seconds to hours ...

The first pressurized-thermal-shock experiment (PTSE-1) in the Heavy-Section Steel Technology (HSST) Program is the most recent of a long successtion of fracture-mechanics experiments that are on a scale that allows important aspects of fracture behavior of reactor pressure vessels to be simulated. Such experiments are the means by which theoretical models of fracture behavior can be evaluated for ...

The temperature response of three generic surface models in each of three locations in an ice condenser containment building were calculated assuming a hydrogen deflagration event and using the HECTR code. The intent of using the three generic surfaces was to conservatively represent surfaces of various types of safety equipment. Analyses were performed for four accident ...

Reported is work completed under the subject contract during the last quarter of 1977, together with a progress report on work still in progress at the end of that period. The goal of the contract has been to improve the understanding of shutdown power in reactors, particularly light water reactors primarily fueled with 235U. Specifically, the aim has been to increase the precision and accuracy of ...

The DC power systems in a nuclear power plant provide control and motive power to valves, instrumentation, emergency diesel generators, and many other components and systems during all phases of plant operation including abnormal shutdowns and accident situations. A specific area of concern is the adequacy of the minimum design requirements for DC power systems, particularly ...

The FRAP-T6 code was developed at the Idaho National Engineering Laboratory (INEL) for the purpose of calculating the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to severe hypothetical loss-of-coolant accidents. An important application of the FRAP-T6 code is to calculate the structural performance ...

Nuclear power plants (NPPs) must ensure that the emergency core cooling system (ECCS) or safety-related containment spray system (CSS) remains capable of performing its design safety function throughout the life of the plant. This requires ensuring that long-term core cooling can be maintained following a postulated loss-of-coolant accident (LOCA). Adequate safety operation ...

Nuclear power plant subcompartment analyses are required to determine the containment pressure distribution that might result from a loss-of-coolant accident. The pressure distribution is used to calculate structural and mechanical design loads. The COMPARE code is used widely to perform subcompartment analysis. However, several simplifying assumptions are utilized to ...

The accounting of containment gas concentration following a loss-of-coolant accident is important in the safety evaluation of hydrogen combustible-gas control systems for nuclear power plants. The COGAP code provides such accounting including the effects of (1) the reaction of zirconium and water; (2) radiolysis of core and sump water; (3) corrosion of zinc, aluminum and ...

Brookhaven National Laboratory is involved in assessing the thermohydraulic models in various advanced codes such as TRAC-PFl and TRAC-BDl. These codes have two fluid formulations which require constitutive relationships describing interfacial mass, energy and momentum transfer. These models are either developed from some separate effect tests or in some cases are ad hoc. There is a need to assess ...

Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such ...

A one-dimensional physical model was developed to study the transient freezing of the molten debris layer (a mixture of UO/sub 2/ fuel and zircaloy cladding) produced in a severe reactivity initiated accident in-pile test and deposited on the inner surface of the test shroud wall. The wall had a finite thickness and was cooled along its outer surface by coolant bypass flow. ...

Some of PWR severe accidents are initiated by loss of all AC power. In these cases, accident would proceed while the primary system pressure is still at high level. Thus it is proposed that an intentional depressurization of the primary system has a poten...

The report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark III type of containment. Ba...

The report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark I type of containment. Base...

Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of ne...

The fission products produced in the reactor core are given in Table 2. The release fractions with a core meltdown and a major rupture of the containment. This release is represented by the BWR-2 release and cesium decrease by a decade from BWR-2 to PWR-4, and from PWR-4 to BEED, while the release fractions

Successful condensate polishing operations maintain control of ionic and particulate impurity transport to the pressurized water reactor (PWR) steam generator and the boiling water reactor (BWR) reactor and recirculation system, thus allowing the units to operate more reliably. This report contains the work presented at EPRI's 2003 Workshop on ...

This report presents proceedings of EPRI's 2001 Plant Chemistry Conference, which brought together approximately 100 industry representatives to discuss experiences and issues regarding nuclear plant chemistry at both pressurized water reactor (PWR) and boiling water reactor (BWR) plants.

Following analysis of its C14 inventory, cladding material from spent PWR and BWR fuel rods was exposed to a saturated NaCl solution for 3 months at 200/sup 0/C in order to gain basic data on C14 behaviour after water ingress into a sealed repository. Cla...

As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective ...

The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and ...

Checkout problems presented include the following: PWR large cold leg break; PWR small cold leg break; PWR intermediate sized cold leg break; BWR large recirculation line break; BWR small recirculation line break; INEL Semiscale small cold leg break; INEL LOFT large cold leg break and INEL ...

Corium-E is a material which could possibly result from a hypothetical light water reactor (LWR) fuel melt accident. It would consist of molten UO/sub 2/ fuel, Zircaloy cladding, stainless steel structural material and lesser amounts of fission products. If molten Corium contacted water in a hypothetical fuel melt accident, it could possibly participate in ...

This investigation provides an assessment of the likelihood and consequences of a severe accident in a spent fuel storage pool - the complete draining of the pool. Potential mechanisms and conditions for failure of the spent fuel, and the subsequent release of the fission products, are identified. Two older PWR and BWR spent fuel ...

A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident ...

Core meltdown accidents are being analyzed to develop an understanding of the risk associated with such postulated accidents and to evaluate the impact of possible mitigating engineering safety equipment. An integral feature of these analyses is the determination of containment building pressurization as a result of loadings imposed by the energy stored in ...

Post-accident sampling systems (PASS), high range gaseous effluent monitors and sampling systems for particulates and iodine in high concentrations have been reviewed at twenty-one licensee sites in Region I of the US Nuclear Regulatory Commission which includes fifteen BWR's and fourteen PWR's. Although most of the ...

Accidents involving fuel melting and eventual contact between the high temperature melt and structural concrete may be hypothesized for both light water thermal reactors and liquid metal cooled breeder reactors. Though these hypothesized accidents have a quite low probability of occurring, it is necessary to investigate the probable natures of the ...

It is axiomatic that the severity of a nuclear reactor accident is determined by the extent of radioactivity escape which results. The main focus of site safety analyses is thus on fission product release and transport. Of all the processes involved, fission product escape from the fuel-cladding region into the primary coolant circuit is perhaps the most simple to describe; ...

Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which devel...

This study examines results of analyses performed with the Source Term Code Package to develop updated source terms using NUREG-0956 methods. The updated source terms are to be used to assess the adequacy of current regulatory source terms used as the basis for equipment qualification. Time-dependent locational distributions of radionuclides within a containment following a severe ...

As a result of the Chernobyl accident and other precursor events (e.g., Diablo Canyon), the US Nuclear Regulatory Commission's (NRC's) Office of Nuclear Regulatory Research (RES) initiated an extensive project during 1989 to carefully examine the potential risks during Low Power and Shutdown (LP S) operations. Shortly after the program began, an event ...

As a result of the Chernobyl accident and other precursor events (e.g., Diablo Canyon), the US Nuclear Regulatory Commission`s (NRC`s) Office of Nuclear Regulatory Research (RES) initiated an extensive project during 1989 to carefully examine the potential risks during Low Power and Shutdown (LP&S) operations. Shortly after the program began, an event occurred at the ...

Design criteria and alternate decay heat removal system concepts which have evolved in several different countries throughout the world were compared. The conclusion was reached that the best way to improve the reliability of pressurized water reactor (PWR) decay heat removal is first to focus on improving the reliability of the auxiliary feedwater and high pressure injection ...

A boiling water reactor (BWR) with a steel containment shell that adopts the passive steel containment system (PSCS) as a part of the residual heat removal system (RHR) is considered as a future BWR configuration. In addition to the usual water injection systems and RHR systems, the future BWR has an overall safety system with more ...

The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. Under primary sponsorship of the US Nuclear Regulatory Commission (NRC), Idaho National Engineering and Environmental Laboratory (INEEL) is responsible for overall maintenance of this code and for improvements for pressurized water reactor ...

A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has ...

This study is an evaluation of a pressurized-water reactor (PWR) accident as defined by WASH 1400 for the proposed nuclear reactor site at Cementon, N. Y. Using an extension of the Environmental Protection Agency's AIREM computer code, the following were ...

The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16...

The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel...

... COMPUTERIZED SIMULATION, *RADIATION DOSAGE, HUMANS, NEW YORK, THESES, IODINE, EXPOSURE(PHYSIOLOGY), RADIOACTIVE ...

A series of four experiments has been conducted at the Argonne National Laboratory in the Transient Reactor Test Facility (TREAT) aimed at providing data for characterizing the radiological source term for severe light water reactor accidents having the greatest contribution to risk (1,2). In each of the tests, four fuel elements were exposed to environmental conditions ...

The LOCA (loss of coolant accident) is a hypothesized, low-probability accident used as a licensing basis for nuclear power plants. Computer codes which have been under development for at least a decade have been the principal tools used to assess the consequences of the hypothesized LOCA. Models exist in two versions. In EM's (Evaluation Models) ...

This report discusses work done in support of the resolution of Generic Issue-82, ''Beyond Design Basis Accidents in Spent Fuel Pools''. Specifically the probability of spent fuel pool failure due to earthquakes was determined for the pools at the Vermont Yankee Nuclear Power Station (BWR) and the H. B. Robinson S.E. ...

The Severe Accident Sequence Analysis (SASA) Program, sponsored by the US Nuclear Regulatory Commission (NRC), is addressing a number of accident scenarios that potentially pose a health hazard to the public. Two of the scenarios being analyzed in detail at the Idaho National Engineering Laboratory (INEL) are the station blackout at the Bellefonte nuclear ...

RETRAN represents a new computer code approach for analyzing the thermal-hydraulic response of Nuclear Steam Supply Systems (NSSS) to hypothetical Loss of Coolant Accidents (LOCA) and Operational Transients. In contrast to the ''conservative'' approach, RETRAN provides ''best estimate'' solutions to ...

RETRAN represents a new computer code approach for analyzing the thermal-hydraulic responses of Nuclear Steam Supply Systems to hypothetical Loss-of-Coolant-Accidents (LOCAs) and operational transients. In contrast to the conservative approach, RETRAN provides best estimate solutions to hypothetical LOCAs and operational transients. RETRAN is a computer code package developed ...

A project is being conducted in the Nuclear Safety Pilot Plant (NSPP), located at the Oak Ridge National Laboratory (ORNL), to study the behavior of aerosols assumed to be generated during LWR reactor accident sequences and released into containment. This project, which is part of the ORNL Aerosol Release and Transport (ART) Program, is sponsored by the Nuclear Regulatory ...

The characterization and dynamic behavior of fine particles are the main subjects of an ongoing investigation of the particle collection effectiveness of the engineered safety feature (ESF) systems in nuclear power plants. This investigation is part of a larger study of the release of radionuclides to the environment from such plants during postulated accidents that are severe ...

Deformation behavior of Zircaloy-4 cladding under simulated loss-of-coolant accident (LOCA) conditions is being investigated in the Multirod Burst Test (MRBT) program in single rod and multirod tests. In these tests, internally-pressurized unirradiated Zircaloy-4 tubes containing internal electrical heaters are heated to failure in a low-pressure, superheated-steam environment ...

The report describes the results of a series of calculations conducted to investigate the response of BWR Mark III containments to short-term station blackout severe accident sequences. The BWR-LTAS, BWRSAR, and MELCOR codes were employed to conduct quant...

The report describes the results of a series of calculations conducted to investigate the response of BWR Mark II containments to short-term and long-term station blackout severe accident sequences. The BWR-LTAS, BWRSAR, and MELCOR codes were employed to ...

A state-of-the-art review of improvements including filter/vent of BWR pressure suppression and PWR ice containments has been carried out. It includes a summary description of operating and planned BWR pressure suppression and PWR ice containments; a review of improvements proposed to-date; an assessment and an ...

A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been ...

The Modular Accident Analysis Program (MAAP) simulates LWR system response to a severe core accident. Overall, calculations performed with the PWR version of MAAP have compared well with a wide variety of other data. These results have proven MAAP an acceptable tool to support individual plant examinations and ...

The estimated costs for post-accident cleanup at the reference BWR (developed previously in NUREG/CR-2601, Technology, Safety and Costs of Decommissioning Reference Light Water Reactors Following Postulated Accidents) are updated to January 1989 dollars i...

The document describes the IAEA Programme on Nuclear Reactor's Accident Management, and analyses the safety assessment of the PWR type reactor. (Atomindex citation 24:067123)

As part of the Severe Accident Sequence Analysis (SASA) program of the Nuclear Regulatory Commission (NRC), threats to containment and ways in which operators can help preserve containment integrity during severe accidents are examined. Severe accidents a...

This paper presents recent activity on control rod drop accident (CRDA) analyses in a typical BWR core and uses a best-estimated kinetics code, the kinetics version of SIMULATE (SIMULATE-K).

The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annula...

The purpose of the ROSA-III experiment with a scaled BWR test facility is to examine primary coolant thermal-hydraulic behavior and performance of ECCS during a postulated loss-of-coolant accident of BWR. The results provide information for verification a...

The purpose of the ROSA-III experiment with a scaled BWR test facility is to examine primary coolant thermalhydraulic behavior and performance during a postulated loss-of-coolant accident of BWR. The results provide information for verification and improv...

BWR-LACP has been a versatile tool for the ORNL SASA program. The development effort was minimal, and the code is fast running and economical. Operator actions are easily simulated and the complete scope of both reactor vessel and primary containment are ...

A common understanding and interpretation of BWR system response and the controlling phenomena in LOCA transients has been achieved through the evaluation and comparison of counterpart tests performed in the ROSA-III and FIST test facilities. These facili...

This report presents a preliminary analysis of fission product revaporization in the Reactor Cooling System (RCS) after the vessel failure. The station blackout transient for BWR Mark I Power Plant is considered. The TRAPMELT3 models of evaporization, che...

Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident co...

This report presents an assessment of the aging (time-dependent degradation) of selected major light water reactor components and structures. The stressors, possible degradation sites and mechanisms, potential failure modes and currently used nondestructive examinations, in-service inspection (ISI) and life assessment methods are discussed for seven major light water reactor components: ...

The Modular Modeling System (MMS) is a computer code developed by EPRI to facilitate modeling the dynamics of fossil-fueled and nuclear steam electric power plants. It is intended to assist in the power plant design process and during later operation for troubleshooting, setting control system gains, validation of simulators, checking operating procedures, scoping of nuclear plant ...

The idea of using casks for interim storage of spent fuel arose at GNS after a very controversial political discussion in 1978, when total passive safety features (including aircraft crash conditions) were required for an above ground spent fuel storage facility. In the meantime, GNS has loaded more than 1000 casks at 25 different storage sites in Germany. GNS cask technology is used in 13 ...

Fuel behavior studies under simulated reactivity-initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor (NSRR) since 1975. This report gives the results of experiments performed from April, 1989 through March, 1990 and discussions of them. A total of 41 tests were carried out during this period. The tests are distinguished into ...

This study developed a set of dominant accident sequences that could be applied generically to domestic commercial PWRs as a standardized basis for a probabilistic seismic risk assessment. This was accomplished by ranking the Zion 1 accident sequences. The pertinent PWR safety systems were compared on a plant-by-plant basis to ...

Operating data on all commercial nuclear power plants of the PWR, HWR, BWR and GCR types in the Western World are analysed statistically to determine whether the explanatory variables size, year of operation, vintage and reactor supplier are significant i...

This report describes the results of non-destructive and laboratory radionuclide measurements, as well as waste classification assessments, of BWR and PWR spent control rod assemblies. The radionuclide inventories of these spent control rods were determin...

Progress is summarized for the following research programs: fission product release from LWR fuel; separate effects program; and Fission Product Transport Test Facility.

An overview of fuel failure rate in Japanese BWRs and PWRs for the period 1964-1980 is presented. The following main causes for PWR and BWR fuel failures are listed: hybridizing, fuel densification, collapse of fuel cladding, fuel rod bowing, fretting. de...

Following a brief historical introduction the concept 'load factor' is defined. The average load factors for European and American BWR and PWR plants and for Swedish plants for the years 1972 - 1976 are given. The figures for the Swedish plants are discus...

An interface which transforms fuel rod heat transfer experimental data from an ORINC plot file into the format of the input tape of a local version of COBRA IV is described as implemented in the program OTOCI. The operation, including input and output, is described.

Topics discussed include: steam electric plants; BWR type reactors; PWR type reactors; thermal efficiency of light water reactors; other types of nuclear power plants; the fission process and nuclear fuel; fission products and reactor afterheat; and reactor safety.

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR...

LWRARC calculates spent fuel decay heat generation rates for standard pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assembly designs for cooling times between 1 and 110 years.

The interfacial drag, droplet entrainment, droplet deposition and droplet-size distributions are important for detailed mechanistic modeling of annular dispersed two-phase flow. In view of this, recently developed correlations for these parameters are pre...

A standard format for the safety analysis report requirements for BWR and PWR cooling system materials and inservice inspection is presented. Information needed by the Regulatory staff in order to conduct its safety evaluations is identified.

The technical and economical aspects arising in turning off LWR nuclear power of 900-1300MW are dealt with, account being taken of the differences between BWR and PWR type reactors. The possible decommissioning proposals, and the disposal or confinememt o...

A method is presented which is acceptable to the NRC staff for implementing quality standards and records requirements with respect to the internals of light-water-cooled reactors during preoperational and initial startup testing.

The regulatory guide lists those Section III ASME Code Cases oriented to reactor coolant pressure boundary component materials and testing that are generally acceptable to the NRC staff for implementation in the licensing of light-water-cooled nuclear power plants.

The regulatory guide lists those Section III ASME Code Cases oriented to reactor coolant pressure boundary component design and fabrication that are generally acceptable to the NRC staff for implementation in the licensing of light-water-cooled nuclear power plants.

A listing is given of those Section 3 ASME code cases oriented to materials and testing that are generally acceptable to the NRC staff for implementation in the licensing of light-water cooled nuclear power plants.

This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant pumps important to license renewal. The intent of this AMG is to...

This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in BWR and PWR commercial nuclear power plant electrical switchgear important to license renewal. The latent of...

Under the terms of a contract with the European Community, a sensitivity study about cross-sections of heavy plutonium isotopes and americium isotopes was carried out for PWR and BWR reactor configurations on the use of recycled plutonium. The main result...