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 ...

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...

It has been demanded to EDF to demonstrate that the probability to induce unacceptable consequences is less than 10/sup -7/ /reactor/year. To meet this safety goal, it was necessary to evaluate the consequences of the loss of redundant systems and more ge...

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 experimental study concerns the consequences on the environment of a PWR severe accident. A preliminary bibliographical survey has been undertaken in order to determine the elements to study, and the experimental protocols to use. 4 fission products ...

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.

The Surry low power and shutdown probabilistic risk analysis (PRA) is an ongoing project at Brookhaven National Laboratory (BNL) to identify and quantify potential accident scenarios that may occur in a pressurized water reactor (PWR) during low power and shutdown. It was initiated as a result of various incidents and accidents that ...

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 primary system of a pressurized water reactor (PWR) operates at a relatively high pressure (15.5 MPa, 2250 psia) and consists of piping and components designed to withstand these pressures. The low-pressure-injection system (LPIS) connects to the primary system but possesses low-pressure piping passing outside the containment. Therefore, a potential exists for a ...

Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are ...

Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to potential nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each protective measure were ...

This report identifies and evaluates accident management strategies that are potentially of value in maintaining containment integrity and controlling the release of radioactivity following a severe accident at a pressurized water reactor with an ice-condenser containment. The strategies are identified using a logic tree structure leading from the safety ...

This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC (Calculations of Reactor ...

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 ...

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...

Information is presented concerning scram failure probability, rod drive failure data, ATWS rule and ATWS requirements, treatment of steam generator tube failures in ATWS evaluation, radiological consequences assessments, ATWS study to include parameter variations and equipment reliability in probabilistic accident analysis, PWR MTC ...

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 16,500 kilograms of depleted natural uranium (as UO/sub 2/), 139 kilograms of ...

This report identifies the results of a program aimed at developing a computer code for use in the analysis of the behavior of iodine during steam generator tube rupture (SGTR) accidents in pressurized water reactors (PWR's). The program was directed towards the identification of the several processes that play a role in the transport and ...

While it would be very difficult to predetermine all of the actions required to mitigate the consequences of every potential severe accident for a nuclear power plant, development of additional guidance and training could improve the likelihood that the operating crew would implement effective sever-accident management measures. The US ...

Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during accident situations. A simple plant ...

The Loss-of-Fluid Test (LOFT) Facility is a scale model of a commercial PWR and is as fully functional and operational as the generic commercial counterpart. LOFT was designed and built for experimental purposes as part of the overall NRC reactor safety research program. The purpose of LOFT is to assess the capability of reactor safety systems to perform their intended ...

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 plant designs. Using ...

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 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...

Accident off-site consequence calculations and risk assessments performed for the ''risk oriented analysis'' of the German prototype fast breeder reactor SNR-300 were performed with a modified version of the off-site accident consequence model UFOMOD. The modifications mainly ...

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

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 for transport of ...

Response of a PWR core loaded with Combined Non-Fertile and UO{sub 2} (CONFU) fuel assemblies to control rod ejection accident was evaluated. A number of core arrangements and TRU fuel compositions were considered and the results were compared with the performance of a reference all-UO{sub 2} core. The comparison was based on the results of a simple point ...

This report identifies and evaluates accident management strategies that are potentially of value in maintaining containment integrity and controlling the release of radioactivity following a severe accident as a pressurized water reactor with large-dry containment. The strategies are identified using a logic tree structure leading from the safety ...

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 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)

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 ...

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 ...

A method is developed for deriving a set of equations relating the public risk in potential nuclear reactor accidents to the basic variables, such as population distributions and radioactive releases, which determine the consequences of the accidents. The equations can be used to determine the risk for different values of the basic ...

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...

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 ...

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 ...

about homeland security research. Browse these EPA Emergency Response subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

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 ...

Detailed three-dimensional MEKIN-B calculations of the PWR control rod ejection accident (REA) are being performed as part of the BNL/NRC evaluation of methods currently used to analyze PWR REA events. A principal objective of these calculations has been to evaluate in three dimensions the effect of flux redistribution on the core ...

The report describes the two-dimensional, Pressurized Water Reactor (PWR) version of the MELPROG computer code, MELPROG/PWR-MOD1. The purpose of MELPROG is to provide for an accident sequence a description of (1) the state of the reactor core and surround...

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. ...

Input models for the Transient Reactor Analysis Code (TRAC) are described and applications of these models to reactor transients involving small breaks in the primary coolant pressure boundary are demonstrated. The operation of the primary overpressure protection system (relief and safety valves) and the thermal-hydraulic response of the reactor to these transients are obtained from numerical ...

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 lradiological a consequences of the Chernobyl accident Proceedings of the first international conference

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.

French studies have early considered the fact that, despite all the precautions taken, the possibility of severe accidents cannot be absolutely excluded; these accidents include core meltdown and a more or less significant loss, at an early or later stage...

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)

Analyses of fuel behavior during LOCA have generally been performed for a single fuel rod (the ''hot rod'') of a reactor core. This type of analysis, even using best estimate assumptions, does not apply to the behavior of majority of the fuel rods in the core. Furthermore, burnup effects are normally considered only so far as the fill gas pressure and, partly, ...

The concern raised by Generic Safety Issue (GSI-191), 'Assessment of Debris Accumulation on PWR Sump Performance,' is the transport of debris to pressurized-water-reactor (PWR) sump screens following a loss-of-coolant accident (LOCA) and subsequent impact...

The study of the risk associated with operation of a PWR includes assessment of severe accidents in which a combination of faults results in melting of the core. Probabilistic methods are used in such assessment, hence it is necessary to estimate the prob...

The overall objective of the FLECHT test program has been to obtain heat transfer data useful for calculating the reflooding behavior of a PWR core following a postulated loss of coolant accident. The behavior of the emergency core coolant (ECC) with resp...

Generic Safety Issue (GSI)-191 Assessment of Debris Accumulation on PWR Sump Performance raised the concern of debris transport to pressurized-water-reactor (PWR) sump screens following a loss-of-coolant accident (LOCA) and subsequent impact to emergency ...

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 ...

In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation from nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays ...

One year ago, the rupture of a feed-water-pipe section just upstream of the economizer resulted in a fatal accident at a US utility drum-boiler unit. The direct cause of the accident was thinning of the pipe wall, apparently the result of erosion/corrosion. An accident similar in origin and consequences occurred in ...

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.

Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occ...

The fundamental objective of the accident management program is to assure, in the event of a severe accident at a nuclear plant, that the effectiveness of personnel and equipment is maximized in preventing or mitigating the consequences of the accident. T...

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 ...

The use of RELAP4/MOD6 computer program for analyzing loss-of-coolant accident conditions in PWR type reactors is described.

The use of RELAP4/MOD6 computer program for analyzing loss-of-coolant accident conditions in PWR type reactors is described. (ERA citation 05:004519)

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 PWR system examined in ...

Plant features and operator actions which have been found to be important in either preventing and mitigating severe accidents in PWRs with ice-condenser containments have been identified. Thus features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Sequoyah plant and from assessments of other relevant studies. ...

The purpose of this report is to estimate the consequences of a Hanford reactor accident with emphasis on the effects at distant points. The potential effects in Canada are estimated as well as the consequences within the United States.

An exploratory sensitivity study with the MELCOR Accident Consequence Code System (MACCS) is presented. This study was performed to provide (1) an indication of the possible impact of consequence modeling uncertainties on the results of an integrated prob...

Experimental studies using a protoypical flow model of a pressurized water reactor (PWR) demonstrate water hammer in the cold legs due to the admission of emergency core cooling (ECC). Such water hammer can occur in an actual PWR during reflood provided t...

Simulating the complicated physical processes that occur in a PWR pressurizer during a transient presented a considerable challenge to modelers. The computer code developed and validated in this study will help utilities to better understand both the behavior of the pressurizer and the overall performance of a PWR after a loss-of-coolant ...

This document presents the preliminary internal events Level 1 results (including fire and flood) obtained as a result of a coarse screening analysis on the low power and shutdown accident frequencies of the Surry Nuclear Power Plant. The work was perform...

For Zion, the following accident sequences were examined: station blackout with failure of auxiliary feedwater (TMLB'), loss of collant accidents (LOCA's) of various sizes inside containment, and the V-sequence interacing system LOCA which involves direct...

Operator Action Event Trees for transient and LOCA initiated accident sequences at the Zion 1 PWR have been developed and documented. These trees logically and systematically portray the role of the operator throughout the progression of the accident. The...

In JAERI, a reactor accident diagnostic system using a method of knowledge engineering has been developed to identify the cause and type of an abnormal transient of a reactor plant. This system consists of a knowledge base and the inference engine named I...

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...

This report is a first documentation of the KfK/PNS activities and plans to investigate the behaviour of steel containments under accident conditions. The investigations will deal with a free standing spherical containment shell built for the latest type ...

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 ...

The report identifies and evaluates accident management strategies that are potentially of value in maintaining containment integrity and controlling the release of radioactivity following a severe accident at a pressurized water reactor with an ice-conde...

This report provides an initial assessment of the impact of proposed Regulatory Guide 1.97 on a typical Westinghouse PWR. The accident monitoring instrumentation requirements and the degree of conformance of existing instrumentation are evaluated. Emphasi...

After the TMI (Three Mile Island) accident, there has been significant interest in analyzing and understanding the phenomena that may occur in a PWR (Pressurized Water Reactor) accident which may lead to partial or total core meltdown and degradation. Natural convection is one of the important phenomena. In the present paper the ...

A computational investigation is undertaken into the role of buoyancy in a PWR boron dilution transient following a postulated Small Break Loss of Coolant Accident (SB-LOCA). In the scenario envisaged there is flow of de-borated and relatively high temperature water from a single cold leg into the downcomer; flow rates are typical of natural circulation ...

This report describes methods for constructing detailed logic sequences that relate individual faults along an entire accident chain. The methods allow for detailed evaluation of system interactions with explicit consideration of mutually exclusive, dependent and common mode events included. A mathematical development of the AND-OR-NOT logic operations and their incorporation ...

This paper discusses the role the Loss-of-Fluid Test (LOFT) Experimental Program has played and will play in addressing licensing issues of interest to the US Nuclear Regulatory Commission (NRC), nuclear steam supply system vendors, and utility power companies. The LOFT facility is an operating, prototypically scaled pressurized water reactor (PWR) system in which experiments ...

Under the loss of heat sinks ATWS (anticipated transients without scram) accident of the PWR operation of nuclear heating reactors with natural circulation, the reactor is probably to enter a low quality two-phase flow density wave instability region and ...

This paper is the final report of the ROSA-II experimental program, in which summary of the integral test results on thermal hydraulic behavior in a loss-of-coolant accident (LOCA) of pressurized water reactor (PWR) and on the effect of emergency core coo...

Supplemental to WAPD-SC-54l. A summary is presented of generalized data on the flammability of hydrogen in steam-air mixtures. The flammability data are applied to a postulated Pressurized Water Reactor loss-of-coolant accident and to estimates of the potential pressure effects on the PWR plant container. (C.H.)

French safety philosophy as concerns severe PWR accidents has already been outlined by the Director of CEA/IPSN in an article published in ''Nuclear Safety''. Therefore the present paper will focus on: (a) the French reference source terms, as used for el...

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 ...

The decision to use mixed-oxide (MOX) fuel in PWR's involved re-investigation of a certain number of accidents and notably control rod ejection transients. It has thus been shown that this accident would be no more severe than in the case of all-uranium c...

The main purpose of this work is to analyse the pressure and temperature behaviour of the metalic containment of a PWR building, submitted to a postulated loss-of-coolant accident (LOCA) caused by a double-ended rupture in the main line of the primary cir...

A combination approach of an expert system and neural networks is used to implement a prototype severe accident diagnostic system which would monitor the progression of the severe accident and provide necessary plant status information to assist the plant staff in accident management during the accident. The ...

In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate ...

Jan 11, 2011 ... trajectories and potential reentry accidents and environments. .... Environmental Consequences of Potential Accidents ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was developed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was completed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The development of two new probabilistic accident consequence codes, MACCS and COSYMA, was developed in 1990. These codes estimate the consequence from the accidental releases of radiological material from hypothesized accidents at nuclear installations. ...

The MELCOR Accident Consequence Code System (MACCS) has been applied to the radiological consequence assessment of potential accidents from a non-reactor nuclear facility. MACCS has been used in a variety of applications to evaluate radiological dose and ...

This report contains technical information used to determine accident consequences for the Spent Nuclear Fuel Project safety documents. It does not determine accident consequences or describe specific accident scenarios, but instead provides generic information.

...mitigate the consequences of an accident; (ii) The equipment...relied on to prevent potential accidents or mitigate their consequences...relied on to prevent potential accidents or mitigate their consequences...evaluated in the Integrated Safety Analysis. (d) The ...

A light water reactor (LWR) fuel assembly design consisting of a blend of weapons-grade plutonium and natural thorium oxides was examined. The design meets current thermal-hydraulic and safety criteria. Such an assembly would have enough reactivity to achieve three cycles of operation. The pin power distribution indicates a fairly level distribution across the assembly, avoiding hot spots near ...

An important consideration in the development and assessment of severe accident management strategies is that while the strategies are often built on the knowledge base of Probabilistic Safety Analyses (PSA), they must be interpretable and meaningful in terms of the control room indicators. In the following, the relationships between PSA and severe ...

Effective planning for the management of severe accidents at nuclear power plants can produce both a reduction in the frequency of such accidents as well as the ability to mitigate their consequences if and when they should occur. This report provides an ...

After some indications on the Chernobylsk power plant and the characteristics of the RBMK series, the scenario of the accident is developed. The immediate consequences and the control of the accident are commented before concluding on more general lessons...

The hypothetical core disruptive accident in an LMFBR is discussed under the following main headings: reactor dynamics; mechanical consequences; and post- accident heat removal. 79 references. (DCC)

The impact on society of the Chernobyl accidents is assessed. The situation prior to Chernobyl with respect to regulations of radiation protection against the consequences of a major accident is considered. The development of the recommendations and regul...

In this report information about the nuclear accident at Chernobyl and the radioactivity burdening of Greece from the radioactive releases of the accident are presented. The main characteristics of the RBMK-1000 reactor and the flow pattern of the radioac...

The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by the Secretary of State for the Environment. The event at the accident site, the release and dispersal of radioactive substances into the atm...

It summarizes the consequences in Sweden of the Chernobyl accident, describes the emergency response, the basis for decisions and countermeasures, the measurement strategies, the activity levels and doses and countermeasures and action levels used. Past a...

Table Top Exercises are used for the training of emergency response personnel from a wide range of disciplines whose duties range from strategic to tactical, from managerial to operational. The exercise reported in this paper simulates the first two or three hours of an imaginary accident on a generic PWR site (named Seaside or Lakeside depending on its ...

IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena ...

Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based ...

This report presents results of analyses of the environmental releases of fission products (source terms) for severe accident scenarios in a pressurized water reactor with an ice-condenser containment. The analyses were performed to support the Severe Accident Risk Reduction/Risk Rebaselining Program (SARRP) which is being undertaken for the US Nuclear ...

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 ...

Studies of Three Mile Island (TMI) accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC and to provide more complete information for ...

This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study ...

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 ...

This booklet is written to help the untrained accident investigator perform a type C'' accident investigation. It is designed to identify the philosophy of accident investigation, the steps involved, the potential problems in the investigation, and give instruction for completing the DOE 5484 X form. A hypothetical ...

This booklet is written to help the untrained accident investigator perform a type ``C`` accident investigation. It is designed to identify the philosophy of accident investigation, the steps involved, the potential problems in the investigation, and give instruction for completing the DOE 5484 X form. A hypothetical ...

Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. ...

An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting ...

This paper presents results of a sensitivity study for potential recovery actions during a station blackout severe accident in a pressurized water reactor (PWR). The accident progression for each of the recovery actions was calculated by a modified version of the severe-accident integrated analysis code MAAP 3.0B. ...

Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual ...

Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 151, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant ...

Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 103, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant ...

This document quantifies the offsite radiological consequences of the bounding flammable gas accident for comparison with the 25 rem Evaluation Guideline established in DOE-STD-3009, Appendix A. The bounding flammable gas accident is a detonation in a sin...

At the fifth anniversary of the Chernobyl accident the initial situation at that time, the control of the consequences to Austria in the present light, as well as the knowledge gained from the accident and its consequences are described. A final estimate ...

A dynamic food-chain model and program, DYFOM-95, for predicting the radiological consequences of nuclear accident has been developed. Processes caused by accident release and which will make an impact on radionuclide concentration in the edible parts of ...

This report on the accident at the Chernobyl nuclear power plant discusses the medical consequences of the accident to personnel and emergency forces exposed to high levels of radiation. Data is presented on the rdiation doses to populations surrounding t...