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Sample records for accident source term

  1. Severe accident source term reassessment

    This paper summarizes the status of the reassessment of severe reactor accident source terms, which are defined as the quantity, type, and timing of fission product releases from such accidents. Concentration is on the major results and conclusions of analyses with modern methods for both pressurized water reactors (PWRs) and boiling water reactors (BWRs), and the special case of containment bypass. Some distinctions are drawn between analyses for PWRs and BWRs. In general, the more the matter is examined, the consequences, or probability of serious consequences, seem to be less. (author)

  2. Source term formation in CANDU severe accidents

    The paper presents the phenomena involved in the most important CANDU severe accident (LOCA+LOECC, SBO, SGTR, EFF). Fission products are grouped in classes taking into consideration the half time, volatility, chemistry and biological activity. An analysis of the paths on which the release of the fission products to the environment occurs is performed. For each type of CANDU severe accident the process of source term formation, the magnitude and structure of source term and also the timing are presented on the basis of SOPHAEROS, CPA and IODE (modules included in ASTEC code) calculations, completed with literature results. The discussion about the involved sources of uncertainties is also presented taking into account the complexity of phenomena, the great number of parameters and limited availability of experimental data. Some general recommendations are developed in order to use the results in achieving the procedures for protective actions during a reactor accident. (authors)

  3. 10 CFR 50.67 - Accident source term.

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Accident source term. 50.67 Section 50.67 Energy NUCLEAR... Conditions of Licenses and Construction Permits § 50.67 Accident source term. (a) Applicability. The... to January 10, 1997, who seek to revise the current accident source term used in their design...

  4. Source term and radiological consequences of the Chernobyl accident

    This report presents the results of a study of the source term and radiological consequences of the Chernobyl accident. The results two parts. The first part was performed during the first 2 months following the accident and dealt with the evaluation of the source term and an estimate of individual doses in the European countries outside the Soviet Union. The second part was performed after August 25-29, 1986 when the Soviets presented in a IAEA Conference in Vienna detailed information about the accident, including source term and radiological consequences in the Soviet Union. The second part of the study reconfirms the source term evaluated in the first part and in addition deals with the radiological consequences in the Soviet Union. Source term and individual doses are calculated from measured post-accident data, reported by the Soviet Union and European countries, microcomputer program PEAR (Public Exposure from Accident Releases). 22 refs

  5. Revised accident source terms for light-water reactors

    Soffer, L. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  6. MAAP4.0.7 severe accident source term analysis

    The Severe Accident Source Term Analysis performed in support of U.S. EPR design certification was conducted using MAAP4.07. The analysis had three distinct goals: to determine the most limiting scenario from a severe accident stand point and incorporate the annulus, fuel and safeguards buildings into the MAAP4.0.7 base model; to develop and document the Level 2 Probabilistic Risk Assessment (PRA) Source Term Analysis; and to develop the input from the PRA Level 2 output to PRA Level 3. The methods of this analysis will be presented in this paper. (authors)

  7. LWR severe accident source term research in the USA

    Fission product releases to the environment, or source terms, arise as a result of a highly diverse group of phenomena involved in any particular severe accident sequence. Because of the multiplicity of accident sequences that can occur for a given plant as well as the diversity of the, as yet, imperfectly understood severe accident phenomena, it is not surprising that reactor accidents such as, for example, those documented in NUREG-1150 have indicated large uncertainties in source terms which represent a significant contribution to the uncertainty in the absolute value of risk. Because of the difficulty and expense involved in performing prototypic experiments, substantial reliance has been placed on the development and validation of detailed mechanistic computer codes for analyzing severe accident phenomena and the source terms associated with them. This paper discusses the extensive research and other efforts that have taken place over the last decade to address the technical issues which have a bearing on being able to describe quantitatively the source term(s) and its characteristics. It also summarizes our present state of knowledge and points out areas where additional research will add further to our understanding. In this context the paper discusses the information that could be provided by the PHEBUS-FP program and its use to assess severe accident integral evaluation codes such as VICTORIA and CONTAIN. Finally, this paper discusses the United States Nuclear Regulatory Commission 's efforts to revise the licensing source term (TID-14844) and the implications of this revision, especially for siting and design of future power plants. (author)

  8. Dependence of severe accident source terms on containment performance

    The results of the BMI-2104 analyses for the assessment of severe accident fission product source terms from the point of view of the effectiveness of the containment in attenuating the radioactivity that is predicted to be released to it are examined. The containment performance assumptions considered include design leakage, leak-before-break, large threshold failure and failure to isolate. The results are expressed and discussed in terms of the effective containment decontamination factors inferred from these analyses

  9. Spallation Neutron Source Accident Terms for Environmental Impact Statement Input

    This report is about accidents with the potential to release radioactive materials into the environment surrounding the Spallation Neutron Source (SNS). As shown in Chap. 2, the inventories of radioactivity at the SNS are dominated by the target facility. Source terms for a wide range of target facility accidents, from anticipated events to worst-case beyond-design-basis events, are provided in Chaps. 3 and 4. The most important criterion applied to these accident source terms is that they should not underestimate potential release. Therefore, conservative methodology was employed for the release estimates. Although the source terms are very conservative, excessive conservatism has been avoided by basing the releases on physical principles. Since it is envisioned that the SNS facility may eventually (after about 10 years) be expanded and modified to support a 4-MW proton beam operational capability, the source terms estimated in this report are applicable to a 4-MW operating proton beam power unless otherwise specified. This is bounding with regard to the 1-MW facility that will be built and operated initially. See further discussion below in Sect. 1.2

  10. Revised accident source terms and control room habitability

    In April 1992, the NRC staff presented to the Commissioners the draft NUREG open-quotes Revised Accident Source Terms for Light-Water Nuclear Power Plants.close quotes This document is the culmination of more than ten years of NRC-sponsored research and represents the first change in the NRC's position on source terms since TID-14844 was issued in 1962. The purpose of this paper is to investigate the impact of the revised source terms on the current approach to analyzing control room habitability as required by 10 CFR 50. Sample calculations are presented that identify aspects of the model requiring clarification before the implementation of the revised source terms. 6 refs., 4 tabs

  11. Source term modelling in case of nuclear accidents

    The relative isotopic composition of the nuclides released during a nuclear accidents depends strongly on the implied mechanisms in the failure of fuel elements, safety barriers and accident dynamics. Also, the released fraction depends on the volatility degree and the temperature attaint in the reactor core and the fuel elements during the accident, respectively. At regime operation temperature, when the fuel sheaths are failed the noble gases (Xe and Kr isotopes), the extremely volatile and volatile fission products (I isotopes and Cs, Te and Ru, respectively) are released into the reactor primary circuit. As the temperature increases, other isotopes are released too. Two tables are given presenting a classification of the isotopes in groups of boiling and melting point temperatures, respectively. From the radiologic point of view, evaluation of the impact of the contaminant radioactive release requires consideration of several factors, namely: - activity, half-life, chemical form, biological hazard, geometrical size of the radioactive aerosols, etc. The activity of each isotope at the reactor stack or at the external walls of the reactor building is called source term. The isotopic and combined activity in a point of the environment located at a given distance from the source is evaluated by means of dispersion models starting from the source term. An expression of the activity of a given isotope in terms of its reactor core inventory and the parameters of the safety barriers is presented

  12. Quantification of severe accidents source terms of BWR 4 reactor with Mark I containment using source term code package

    Severe accident source terms of a nuclear power plant which employs a BWR4 reactor with a Mark I containment are quantified with the Source Term Code Package (STCP). Accident scenarios selected for source terms analyses are defined based on the Probabilistic Risk Assessment (PRA) results of accident sequence grouping, containment responses, containment phenomenological event trees, and release category analyses of studies. Included in the paper is a brief description of the structure and major features of STCP together with the modifications made to the code package for the present analysis, the plant model adopted for the STCP source terms quantifications; a presentation and discussion of the source terms as predicted by the STCP for the ten accident sequences analyzed. (orig.)

  13. Source term analysis for a nuclear submarine accident

    A source term analysis has been conducted to determine the activity release into the environment as a result of a large-break loss-of-coolant accident aboard a visiting nuclear-powered submarine to a Canadian port. This best-estimate analysis considers the fractional release from the core, and fission product transport in the primary heat transport system, primary containment (i.e. reactor compartment) and submarine hull. Physical removal mechanisms such as vapour and aerosol deposition are treated in the calculation. Since a thermalhydraulic analysis indicated that the integrity of the reactor compartment is maintained, release from the reactor compartment will only occur by leakage; however, it is conservatively assumed that the secondary containment is not isolated for a 24-h period where release occurs through an open hatch in the submarine hull. Consequently, during this period, the activity release into the atmosphere is estimated as 4.6 TBq, leading to a maximum individual dose equivalent of 0.5 mSv at 800 metres from the berthing location. This activity release is comparable to that obtained in the BEREX TSA study (for a similar accident scenario) but is four orders of magnitude less than that reported in the earlier Davis study where, unrealistically, no credit had been taken for the containment system or for any physical removal processes. (author)

  14. Analysis of the source term in the Chernobyl-4 accident

    The report presents the analysis of the Chernobyl accident and of the phenomena with major influence on the source term, including the chemical effects of materials dumped over the reactor, carried out by the Chair of Nuclear Technology at Madrid University under a contract with the CEC. It also includes the comparison of the ratio (Cs-137/Cs-134) between measurements performed by Soviet authorities and countries belonging to the Community and OECD area. Chapter II contains a summary of both isotope measurements (Cs-134 and Cs-137), and their ratios, in samples of air, water, soil and agricultural and animal products collected by the Soviets in their report presented in Vienna (1986). Chapter III reports on the inventories of cesium isotopes in the core, while Chapter IV analyses the transient, especially the fuel temperature reached, as a way to deduce the mechanisms which took place in the cesium escape. The cesium source term is analyzed in Chapter V. Normal conditions have been considered, as well as the transient and the post-accidental period, including the effects of deposited materials. The conclusion of this study is that Chernobyl accidental sequence is specific of the RBMK type of reactors, and that in the Western world, basic research on fuel behaviour for reactivity transients has already been carried out

  15. Regulatory impact of nuclear reactor accident source term assumptions. Technical report

    This report addresses the reactor accident source term implications on accident evaluations, regulations and regulatory requirements, engineered safety features, emergency planning, probabilistic risk assessment, and licensing practice. Assessment of the impact of source term modifications and evaluation of the effects in Design Basis Accident analyses, assuming a change of the chemical form of iodine from elemental to cesium iodide, has been provided. Engineered safety features used in current LWR designs are found to be effective for all postulated combinations of iodine source terms under DBA conditions. In terms of potential accident consequences, it is not expected that the difference in chemical form between elemental iodine and cesium iodide would be significant. In order to account for the current information on source terms, a spectrum of accident scenerios is discussed to realistically estimate the source terms resulting from a range of potential accident conditions

  16. Source term estimation during incident response to severe nuclear power plant accidents

    This document presents a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. 39 refs., 48 figs., 19 tabs

  17. Source term estimation during incident response to severe nuclear power plant accidents. Draft

    The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. The goal is to present a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. (author)

  18. Influence of accident management strategies on source terms of VVER-1000-type reactors

    The source term can be mitigated by effective accident management. The goal of this work is the investigation of the influence of a number of accident management strategies on the source term of a VVER-1000-type reactor. This work is one of a series of studies investigating the behavior of a VVER-1000-type reactor during severe accidents. In particular, it is based on the study in which the pressure rise in the containment and the melt-through of the cavity bottom was investigated, indicating potential mitigation strategies. To rate the usefulness of these strategies, the source terms of selected scenarios are also calculated in the present work. All the calculations were performed using the Source Term Code Package; hydrogen explosions are not considered. For the first time, the source term behavior of these scenarios was simulated up to the very end of the accident the solidification of the melt

  19. Project on Transfer Mechanism of Radioactive Source Term Under Severe Accident

    SUN; Xue-ting; JI; Song-tao; CHEN; Lin-lin

    2012-01-01

    <正>The "Transfer mechanism of radioactive source term under severe accident" is a sub-project of the research program of "Mechanism and phenomenology of severe accident". An aerosol transfer mechanism experimental facility is built to simulate the passive containment cooling system (PCCS) of advanced pressurizer reactors to research effects to the transfer process of fission products under severe accident. An advanced CFD method is also utilized to research the effects. The objective of this project is to understand

  20. Review of Past Nuclear Accidents: Source Terms and Recorded Gamma-Ray Spectra

    Sanderson, D.C.W.; Cresswell, A.; Allyson, J.D.; McConville, P.

    1997-01-01

    Airborne gamma ray spectrometry using high volume scintillation detectors, optionally in conjunction with Ge detectors, has potential for making rapid environmental measurements in response to nuclear accidents. A literature search on past nuclear accidents has been conducted to define the source terms which have been experienced so far. Selected gamma ray spectra recorded after past accidents have also been collated to examine the complexity of observed behaviour.

  1. The Chernobyl reactor accident source term: development of a consensus view

    Ten years after the reactor accident at Chernobyl, a great deal more data is available concerning the events, phenomena, and processes that took place. The purpose of this document is to examine what is known about the radioactive materials released during the accident, a task that is substantially more difficult than it might first appear to be. The Chernobyl station, like other nuclear power plants, was not instrumented to characterize a disastrous accident. The accident was peculiar in the sense that radioactive materials were released, at least initially, in an exceptionally energetic plume and were transported far from the reactor site. Release of radioactivity from the plant continued for several days. Characterization of the contamination caused by the releases of radioactivity has had a much lower priority than remediation of the contamination. Consequently, an assessment of the Chernobyl accident source term must rely to a significant extent on inferential evidence. The assessment presented here begins with an examination of the core inventories of radioactive materials. In subsequent sections of the report, the magnitude and timing of the releases of radioactivity are described. Then, the composition, chemical forms, and physical forms of the releases are discussed. A number of more recent publications and results from scientists in Russia and elsewhere have significantly improved the understanding of the Chernobyl source term. Because of the special features of the reactor design and the peculiarities of the Chernobyl accident, the source term for the Chernobyl accident is of limited applicability to the safety analysis of other types of reactors

  2. Advanced sodium fast reactor accident source terms : research needs.

    Powers, Dana Auburn; Clement, Bernard [IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France; Denning, Richard [Ohio State University, Columbus, OH; Ohno, Shuji [Japan Atomic Energy Agency, Ibaraki, Japan; Zeyen, Roland [Institute for Energy Petten, Saint-Paul-lez-Durance, France

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic eventEnergetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolantEntrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached claddingRates of radionuclide leaching from fuel by liquid sodiumSurface enrichment of sodium pools by dissolved and suspended radionuclidesThermal decomposition of sodium iodide in the containment atmosphereReactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  3. The Chernobyl reactor accident source term: Development of a consensus view

    In August 1986, scientists from the former Soviet Union provided the nuclear safety community with an impressively detailed account of what was then known about the Chernobyl accident. This included assessments of the magnitudes, rates, and compositions of radionuclide releases during the ten days following initiation of the accident. A summary report based on the Soviet report, the oral presentations, and the discussions with scientists from various countries was issued by the International Atomic Energy Agency shortly thereafter. Ten years have elapsed since the reactor accident at Chernobyl. A great deal more data is now available concerning the events, phenomena, and processes that took place. The purpose of this document is to examine what is known about the radioactive materials released during the accident. The accident was peculiar in the sense that radioactive materials were released, at least initially, in an exceptionally energetic plume and were transported far from the reactor site. Release of radioactivity from the plant continued for about ten days. A number of more recent publications and results from scientists in Russia and elsewhere have significantly improved our understanding of the Chernobyl source term. Because of the special features of the reactor design and the pecularities of the Chernobyl accident, the source term for the Chernobyl accident is of limited applicability of the safety analysis of other types of reactors

  4. Accident source terms for boiling water reactors with high burnup cores.

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  5. CFD Analysis of Migration Mechanism of Source Term Under Severe Accident

    CHEN; Lin-lin; SUN; Xue-ting; JI; Song-tao

    2013-01-01

    The analysis of the migration of source term under severe accident is one of the important aspects of‘Studies on Migration Mechanism of the Source Term under Severe Accident’,which is a significant task of the National Large Advanced PWR Research Program.This research aims at building up a method for analyzing fission product behavior in the containment with CFD code.The effect of PCCS(Passive

  6. Development of the source term PIRT based on findings during Fukushima Daiichi NPPs accident

    Highlights: • We developed the source term PIRT based on findings during the Fukushima accident. • The FoM is the masses or fractions of radionuclides released into the environment. • 68 phenomena were identified as influencing to the FoM. • Radionuclide release from molten fuel had the highest score in the early phase. • MCCI, iodine chemistry, and chemical form had the highest score in the later phase. - Abstract: Research Expert Committee on Evaluation of Severe Accident of AESJ (Atomic Energy Society of Japan) has developed thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and source term (ST) PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aim to explore the debris distribution and the current condition in the NPPs with high accuracy and to extract higher priority from the aspect of the sophistication of the analytical technology to predict the severe accident phenomena by the analytical codes. The ST PIRT is divided into 3 phases for time domain and 9 categories for spatial domain. The 68 phenomena have been extracted and the importance from the viewpoint of the source term has been ranked through brainstorming and discussions among experts. The present paper describes the developed ST PIRT list and summarizes the high ranked phenomena in each phase

  7. Estimation of Source Term from A hypothetical Accident in ETRR-2 Reactor at Inshas,Egypt

    The off-site consequences of an accident in nuclear facilities depend on the source term and the meteorological conditions prevailing during the accident. The present study concerned with the estimation of the source term for the ETRR-2 research reactor by assumption of a hypothetical accident of type loss of coolant accident (LOCA). Core inventory calculations were carried out by utilizing the Monte Carlo code MULTI-KENO and ORIGEN-JR code. The first code was used to calculate the average neutron flux at the core assuming that the water above the core was lost. The results were used as input to the second code, which gave the isotope radio-activities of the core in curies. A comparative study has been done with the data in safety analysis report (SAR), which shows a good agreement with our calculations. The released fractions of the core inventory (source term) was calculated by multiplying the core inventory with the fraction of fission products liberated from the core to the pool water and the fraction of fission products liberated from water to the atmosphere

  8. Accident source terms for Light-Water Nuclear Power Plants. Final report

    In 1962 tile US Atomic Energy Commission published TID-14844, ''Calculation of Distance Factors for Power and Test Reactors'' which specified a release of fission products from the core to the reactor containment for a postulated accident involving ''substantial meltdown of the core''. This ''source term'', tile basis for tile NRC's Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC's reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ''source term'' release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ''source term'' is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it

  9. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  10. Source term evaluation for the upgraded LEU Pakistan Research Reactor-1 under severe accidents

    Research highlights: → Evaluation of source term was done for an upgraded LEU based PARR-I system with a Matlab based computer program having ORIGEN2 code as subroutine for core inventory calculations. → Various accident scenarios, with instantaneous release of radioactivity to containment, have been considered including the startup, fuel loading, and loss-of-coolant accidents. → The source term and containment retention factor values show a rapid increase followed by an approach towards saturation values as the exhaust rates are increased. → The isotope-dependency of the containment retention factor indicates strong sensitivity for 85Kr, 137Xe, 138Xe and 138Cs towards exhaust rate values. - Abstract: Evaluation of source term has been carried out for the upgraded LEU PARR-I system taken as a typical material test reactor (MTR). The modeling and simulation of release of radioactivity has been carried out by developing a Matlab based computer program which uses the ORIGEN2 code for core inventory calculations. For post 180 full-power days continuous operation, various accident scenarios, with instantaneous release of radioactivity to containment, have been considered including the startup, fuel loading, and loss-of-coolant accidents. For noble gases, iodine and for aerosols, the release rate studies have been carried out for the normal, emergency and for the isolation states of containment. The values of source term as well as that of containment retention factor show rapid increase followed by an approach towards saturation values as the exhaust rate values are increased. The isotope-dependency of the containment retention factor has been studied and the results indicate strong sensitivity for 85Kr, 137Xe, 138Xe and 138Cs towards exhaust rate values.

  11. Source term evaluation for the upgraded LEU Pakistan Research Reactor-1 under severe accidents

    Ullah, Sana [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences (PIEAS), P.O. Nilore, Islamabad 45650 (Pakistan); Awan, Saeed Ehsan [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences (PIEAS), P.O. Nilore, Islamabad 45650 (Pakistan); Mirza, Nasir M., E-mail: nasirmm@yahoo.co [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences (PIEAS), P.O. Nilore, Islamabad 45650 (Pakistan); Mirza, Sikander M. [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences (PIEAS), P.O. Nilore, Islamabad 45650 (Pakistan)

    2010-11-15

    Research highlights: {yields} Evaluation of source term was done for an upgraded LEU based PARR-I system with a Matlab based computer program having ORIGEN2 code as subroutine for core inventory calculations. {yields} Various accident scenarios, with instantaneous release of radioactivity to containment, have been considered including the startup, fuel loading, and loss-of-coolant accidents. {yields} The source term and containment retention factor values show a rapid increase followed by an approach towards saturation values as the exhaust rates are increased. {yields} The isotope-dependency of the containment retention factor indicates strong sensitivity for {sup 85}Kr, {sup 137}Xe, {sup 138}Xe and {sup 138}Cs towards exhaust rate values. - Abstract: Evaluation of source term has been carried out for the upgraded LEU PARR-I system taken as a typical material test reactor (MTR). The modeling and simulation of release of radioactivity has been carried out by developing a Matlab based computer program which uses the ORIGEN2 code for core inventory calculations. For post 180 full-power days continuous operation, various accident scenarios, with instantaneous release of radioactivity to containment, have been considered including the startup, fuel loading, and loss-of-coolant accidents. For noble gases, iodine and for aerosols, the release rate studies have been carried out for the normal, emergency and for the isolation states of containment. The values of source term as well as that of containment retention factor show rapid increase followed by an approach towards saturation values as the exhaust rate values are increased. The isotope-dependency of the containment retention factor has been studied and the results indicate strong sensitivity for {sup 85}Kr, {sup 137}Xe, {sup 138}Xe and {sup 138}Cs towards exhaust rate values.

  12. Revaporisation of fission product deposits in the primary circuit and its impact on accident source term

    BOTTOMLEY Paul; KNEBEL KEVIN; VAN WINCKEL Stefaan; HASTE Tim; Souvi, Sidi,; AUVINEN Ari; KALILAINEN J.; KÄRKELÄ Teemu

    2014-01-01

    Chemical revaporisation or physical resuspension of fission product deposits from the primary circuit is now recognised to be a major source term in the late phase of severe fuel degradation in a severe nuclear accident. These results come from tests carried out under different experimental projects in the European Commission (EC) Framework Programmes. These include the revaporisation tests carried out at the Transuranium Institute (ITU), Karlsruhe under the Fourth Framework Programme, the Ph...

  13. SARNET: Integrating Severe Accident Research in Europe - Safety Issues in the Source Term Area

    SARNET (Severe Accident Research Network) is a Network of Excellence of the EU 6. Framework Programme that integrates in a sustainable manner the research capabilities of about fifty European organisations to resolve important remaining uncertainties and safety issues concerning existing and future nuclear plant, especially water-cooled reactors, under hypothetical severe accident conditions. It emphasises integrating activities, spreading of excellence (including knowledge transfer) and jointly-executed research. This paper summarises the main results obtained at the middle of the current 4-year term, highlighting those concerning radioactive release to the environment. Integration is pursued through different methods: the ASTEC integral computer code for severe accident modelling, development of PSA level 2 methods, a means for definition, updating and resolution of safety issues, and development of a web database for storing experimental results. These activities are helped by an evolving Advanced Communication Tool, easing communication amongst partners. Concerning spreading of excellence, educational courses covering severe accident analysis methodology and level 2 PSA have been organised for early 2006. A text book on Severe Accident Phenomenology is being written. A mobility programme for students and young researchers has started. Results are disseminated mainly through open conference proceedings, with journal publications planned. The 1. European Review Meeting on Severe Accidents in November 2005 covered SARNET activities during its first 18 months. Jointly executed research activities concern key issues grouped in the Corium, Containment and Source Term areas. In Source Term, behaviour of the highly radio-toxic ruthenium under oxidising conditions, including air ingress, is investigated. Models are proposed for fuel and ruthenium oxidation. Experiments on transport of oxide ruthenium species are performed. Reactor scenario studies assist in defining

  14. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  15. Inversion method of source term in nuclear accident based on Gaussian puff model

    The inverse problem of source terms information estimation in nuclear accident is important for emergency response. In this study a review of data assimilation applied on atmospheric dispersion is given. For the atmospheric dispersion model is nonlinear and with model errors, ensemble Kalman filter is adopted for data assimilation. The dispersion consequences is described by Gaussian puff model, and the source term emission rate and release height is estimated real-time. To determine the best first guess parameters' value and errors, more than 10 twin experiments have been carried on. The results show that the ensemble Kalman filter can be applied successfully to estimate the source term information when there are one or two unknown parameters, the estimated accuracy is related to first guess value, and is impacted by the standard deviation of perturbation. To reduce the estimation error, first guess value setting to the half to two times of true value is recommended. (author)

  16. Source term and behavioural parameters for a postulated HIFAR loss-of-coolant accident

    The fraction of the fission product inventory which might be released into the atmosphere of the HIFAR reactor containment building (RCB) during a postulated loss-of-coolant accident (LOCA) has been evaluated as a function of time, for each classification of airborne radioactivity. This appraisal will be used as the source term for a computer program, which uses realistic attenuation of the fission product aerosol in a single compartment model with a defined leakrate to predict possible radioactive releases into the environment in a hypothetical bounding case reactor accident which is rather more severe in all major aspects than any single LOCA. Also given are the parameters governing the attenuation of the aerosol and vapours in the atmosphere of the RCB so that their behaviour may be accurately modelled. The source terms for several other types of accident involving the meltdown of fuel elements have also been considered but in less detail than the LOCA case. In some of the cases, the fission products are released directly to atmosphere, so there is no attenuation of the release by deposition within the RCB

  17. Severe accident source terms for a sodium-cooled fast reactor

    Highlights: • This study analyzes offsite doses for characteristic SFR scenarios. • Models to calculate the source term for an SFR were developed for this work. • Environmental releases are small due to effectiveness of retention mechanisms. • NRC’s Quantitative Health Objectives are satisfied with high margins. - Abstract: In order to support the demonstration of a risk-informed approach to the design optimization of a sodium-cooled fast reactor (SFR), it was necessary to make realistic estimates of the consequences of severe accident scenarios. This paper describes the database, models, and assumptions used to estimate the offsite consequences of characteristic severe accident scenarios. As required for comparison with the NRC’s technology neutral framework limit curve, the offsite dose at one mile from the plant boundary is calculated using conservative meteorology. The reference plant design is a 1000 MWt pool-type design with metallic fuel. Because an integrated analysis tool comparable to MELCOR does not exist for SFR accident scenario analysis, it was necessary to write a computer code that would assess release of radionuclides from the fuel and transport within the reactor primary system and to link those analyses with results from existing computer codes that assess the dynamic response of the reactor, containment thermal–hydraulics, and radionuclide transport processes within the containment. The analyses indicate that the offsite source terms for SFR severe accident scenarios tend to be small because of the low melting temperature of the fuel, likelihood of significant retention of fission products within the sodium pool, augmentation of containment deposition processes by interaction with sodium oxide aerosols, and small driving force for release from the containment to the environment. A number of major sources of modeling uncertainty are identified as requiring further development effort. An integrated modeling capability, similar to the

  18. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  19. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences

  20. Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ''MELCOR Verification, Benchmarking, and Applications,'' whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR

  1. Source terms analysis of a maximum release accident for an AGN-201M reactor

    The fundamental liability of any nuclear reactor is the possibility of exposing the public and environment to an excessive level of nuclear radiation. In a previous paper, the authors addressed the risk and potential vulnerability assessment of a maximum hypothetical release accident (MHRA) for the AGN-201M reactor at the University of New Mexico. The MHRA is defined as the total release of all radiological effluents from the reactor facility to the environment. A level I probabilistic risk assessment was performed to assess the risk to the public. The type of effluents, total activity, maximum exposure rate, and related health effects associated with an MHRA were analyzed in an attempt to identify the source term and its consequences. The source term was characterized for the worst-case scenario only because the magnitude of the released effluents is deemed ineffectual for any subcategory release

  2. Specific features of cesium chemistry and physics affecting reactor accident source term predictions

    In the process of assessing remaining uncertainties in predicting the source term of severe reactor accidents, a special investigation is devoted in this report to the case of cesium. The cesium isotopes, especially Cs137 and Cs134, are among those nuclides which could have a major impact on the environment in the event of a release. The processes for release from fuel and retention in the reactor coolant system and the containment are presented. Releases to the atmosphere are also discussed. The intention is to identify and discuss those specific features of cesium chemistry and physics that strongly affect source term predictions. The report has been prepared on contract from the Swedish Nuclear Power Inspectorate as a contribution to the cooperative work within international experts groups of OECD/NEA

  3. SARNET. Severe Accident Research Network - key issues in the area of source term

    About fifty European organisations integrate in SARNET (Network of Excellence of the EU 6th Framework Programme) their research capacities in resolve better the most important remaining uncertainties and safety issues concerning existing and future Nuclear Power Plants (NPPs) under hypothetical Severe Accident (SA) conditions. Wishing to maintain a long-lasting cooperation, they conduct three types of activities: integrating activities, spreading of excellence and jointly executed research. This paper summarises the main results obtained by the network after the first year, giving more prominence to those from jointly executed research in the Source Term area. Integrating activities have been performed through different means: the ASTEC integral computer code for severe accident transient modelling, through development of PSA2 methodologies, through the setting of a structure for definition of evolving R and D priorities and through the development of a web-network of data bases that hosts experimental data. Such activities have been facilitated by the development of an Advanced Communication Tool. Concerning spreading of excellence, educational courses covering Severe Accident Analysis Methodology and Level 2 PSA have been set up, to be given in early 2006. A detailed text book on Severe Accident Phenomenology has been designed and agreed amongst SARNET members. A mobility programme for students and young researchers is being developed, some detachments are already completed or in progress, and examples are quoted. Jointly executed research activities concern key issues grouped in the Corium, Containment and Source Term areas. In Source Term, behaviour of the highly radio-toxic ruthenium under oxidising conditions (like air ingress) for HBU and MOX fuel has been investigated. First modelling proposals for ASTEC have been made for oxidation of fuel and of ruthenium. Experiments on transport of highly volatile oxide ruthenium species have been performed. Reactor

  4. Effect of hypoiodous acid volatility on the iodine source term in reactor accidents

    Routamo, T. [Imatran Voima Oy, Vantaa (Finland)

    1996-12-01

    A FORTRAN code ACT WATCH has been developed to establish an improved understanding of essential radionuclide behaviour mechanisms, especially related to iodine chemistry, in reactor accidents. The accident scenarios calculated in this paper are based on the Loss of Coolant accident at the Loviisa Nuclear Power Plant. The effect of different airborne species, especially HIO, on the iodine source term has been studied. The main cause of the high HIO release in the system modelled is the increase of I{sub 2} hydrolysis rate along with the temperature increase, which accelerates HIO production. Due to the high radiation level near the reactor core, I{sub 2} is produced from I{sup -}very rapidly. High temperature in the reactor coolant causes I{sub 2} to be transformed into HIO and through the boiling of the coolant volatile I{sub 2} and HIO are transferred efficiently into the gas phase. High filtration efficiency for particulate iodine causes I{sup -} release to be much lower than those of I{sub 2} and HIO. (author) 15 figs., 1 tab., refs.

  5. Methods to prevent the source term of methyl lodide during a core melt accident

    The purpose of this literature review is to gather available information of the methods to prevent a source term of methyl iodide during a core melt accident. The most widely studied methods for nuclear power plants include the impregnated carbon filters and alkaline additives and sprays. It is indicated that some deficiencies of these methods may emerge. More reactive impregnants and additives could make a great improvement. As a new method in the field of nuclear applications, the potential of transition metals to decompose methyl iodide, is introduced in this review. This area would require an additional research, which could elucidate the remaining questions of the reactions. The ionization of the gaseous methyl iodide by corona-discharge reactors is also shortly described. (au)

  6. A source term estimation method for a nuclear accident using atmospheric dispersion models

    Kim, Minsik; Ohba, Ryohji; Oura, Masamichi;

    2015-01-01

    The objective of this study is to develop an operational source term estimation (STE) method applicable for a nuclear accident like the incident that occurred at the Fukushima Dai-ichi nuclear power station in 2011. The new STE method presented here is based on data from atmospheric dispersion...... models and short-range observational data around the nuclear power plants.The accuracy of this method is validated with data from a wind tunnel study that involved a tracer gas release from a scaled model experiment at Tokai Daini nuclear power station in Japan. We then use the methodology developed...... and validated through the effort described in this manuscript to estimate the release rate of radioactive material from the Fukushima Dai-ichi nuclear power station....

  7. Methods to prevent the source term of methyl lodide during a core melt accident

    Karhu, A. [VTT Energy (Finland)

    1999-11-01

    The purpose of this literature review is to gather available information of the methods to prevent a source term of methyl iodide during a core melt accident. The most widely studied methods for nuclear power plants include the impregnated carbon filters and alkaline additives and sprays. It is indicated that some deficiencies of these methods may emerge. More reactive impregnants and additives could make a great improvement. As a new method in the field of nuclear applications, the potential of transition metals to decompose methyl iodide, is introduced in this review. This area would require an additional research, which could elucidate the remaining questions of the reactions. The ionization of the gaseous methyl iodide by corona-discharge reactors is also shortly described. (au)

  8. A comparison of world-wide uses of severe reactor accident source terms

    The definitions of source terms to reactor containments and source terms to the environment are discussed. A comparison is made between the TID-14844 example source term and the alternative source term described in NUREG-1465. Comparisons of these source terms to the containments and those used in France, Germany, Japan, Sweden, and the United Kingdom are made. Source terms to the environment calculated in NUREG-1500 and WASH-1400 are discussed. Again, these source terms are compared to those now being used in France, Germany, Japan, Sweden, and the United Kingdom. It is concluded that source terms to the containment suggested in NUREG-1465 are not greatly more conservative than those used in other countries. Technical bases for the source terms are similar. The regulatory use of the current understanding of radionuclide behavior varies among countries

  9. An artificial neural network approach to reconstruct the source term of a nuclear accident

    This work makes use of one of the main features of artificial neural networks, which is their ability to 'learn' from sets of known input and output data. Indeed, a trained artificial neural network can be used to make predictions on the input data when the output is known, and this feedback process enables one to reconstruct the source term from field observations. With this aim, an artificial neural networks has been trained, using the projections of a segmented plume atmospheric dispersion model at fixed points, simulating a set of gamma detectors located outside the perimeter of a nuclear facility. The resulting set of artificial neural networks was used to determine the release fraction and rate for each of the noble gases, iodines and particulate fission products that could originate from a nuclear accident. Model projections were made using a large data set consisting of effective release height, release fraction of noble gases, iodines and particulate fission products, atmospheric stability, wind speed and wind direction. The model computed nuclide-specific gamma dose rates. The locations of the detectors were chosen taking into account both building shine and wake effects, and varied in distance between 800 and 1200 m from the reactor.The inputs to the artificial neural networks consisted of the measurements from the detector array, atmospheric stability, wind speed and wind direction; the outputs comprised a set of release fractions and heights. Once trained, the artificial neural networks was used to reconstruct the source term from the detector responses for data sets not used in training. The preliminary results are encouraging and show that the noble gases and particulate fission product release fractions are well determined

  10. Release of radionuclides following severe accident in interim storage facility. Source term determination

    Among the severe accidents that can cause the release of radionuclides from an interim storage facility, with a consequent relevant radiological impact on the population, there is the impact of an aircraft on the facility. In this work, a safety assessment analysis for the case of an aircraft crash into an interim storage facility is tackled. To this aim a methodology, based upon DOE, IAEA and NUREG standard procedures and upon conservative yet realistic hypothesis, has been developed in order to evaluate the total radioactivity, source term, released to the biosphere in consequence of the impact, without recurring to the use of complicated numerical codes. The procedure consists in the identification of the accidental scenarios, in the evaluation of the consequent damage to the building structures and to the waste packages and in the determination of the total release of radionuclides through the building-atmosphere interface. The methodology here developed has been applied to the case of an aircraft crash into an interim storage facility currently under design. Results show that in case of perforation followed by a fire incident the total released activity would be greater of some orders of magnitude with respect to the case of mere perforation. (author)

  11. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m2, fluence 3 MW-y/m2, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass

  12. Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

    Cambi, G. [Bologna Univ. (Italy). Physics Dept.; Cepraga, D.G. [ENEA, Bologna (Italy). Innovation Dept.; Di Pace, L. [ENEA, Frascati (Italy). Fusion Sector CR di Frascati; Porfiri, M.T. [ENEA, Frascati (Italy). Fusion Dept.

    1995-12-31

    The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SB heat transport system HTS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m{sup 2}, fluence 3 MW-y/m{sup 2}, average pulse length 1,000 s, dwell time 1,200 s. The structural material for FW and SB is AISI 316L. (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SB heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass.

  13. Revaporisation of fission product deposits in the primary circuit and its impact on accident source term

    Highlights: • Testing showed all volatile fission product deposits revaporise during reheating. • Cs deposits revaporise in flowing steam above ∼500 °C up to 92% by 1000 °C. • Revaporisation can occur in all atmospheres, contributing to late stage releases. • Separate effect testing of Cs, I and H2O has determined key reactions and species. • Ab initio modelling has identified stable surface–adsorbed compounds. - Abstract: Chemical revaporisation or physical resuspension of fission product deposits from the primary circuit is now recognised to be a major source term in the late phase of fuel degradation in a severe nuclear accident. These results come from tests carried out under different experimental projects in the European Commission (EC) Framework Programmes. These include the revaporisation tests carried out at the Transuranium Institute (ITU), Karlsruhe under the Fourth Framework Programme, the Phébus FP post-test analysis programme that examined FPT1, FPT3 and FPT4 deposits in separate-effect tests as well as EXSI-PC tests carried out at VTT, Espoo. The first tests at ITU and VTT concentrated on the behaviour of caesium as a very important fission product; this has helped detailed interpretation of the integral Phébus FP tests and has clarified some puzzling observations. Testing with Phébus FPT1 and FPT4 deposits at ITU demonstrated that revaporisation is a likely, rather than a possible, phenomenon with a severely degrading bundle. They have also shown that any changes in temperature (substrate or gas), flow rate or atmosphere composition or pressure can lead to the volatilisation or removal of the deposited caesium. Cs was particularly easy to follow given the high activity levels of Cs in the deposit. However further analysis of the deposits shows that other fission products are also subject to revaporisation. In the most recent FPT3 test, the chemical analysis of the filters has enabled examination of other fission products and

  14. Re-examination of the steam explosion source term during severe accidents

    Some analyses of the potential contribution of steam explosions to the radioactive material release during severe reactor accidents are described. Experimental data on debris comminution are discussed and it is suggested that mechanical aerosolization will not be a major source of aerosols during steam explosions. The thermodynamics and kinetics of ruthenium vaporization are analysed. Ruthenium vaporization is shown to be an important contributor to radionuclide release if debris particles produced by steam explosions are small and hot. (author)

  15. Application of Bayesian nonparametric models to the uncertainty and sensitivity analysis of source term in a BWR severe accident

    A full-scope method is constructed to reveal source term uncertainties and to identify influential inputs during a severe accident at a nuclear power plant (NPP). An integrated severe accident code, MELCOR Ver. 1.8.5, is used as a tool to simulate the accident similar to that occurred at Unit 2 of the Fukushima Daiichi NPP. In order to figure out how much radioactive materials are released from the containment to the environment during the accident, Monte Carlo based uncertainty analysis is performed. Generally, in order to evaluate the influence of uncertain inputs on the output, a large number of code runs are required in the global sensitivity analysis. To avoid the laborious computational cost for the global sensitivity analysis via MELCOR, a surrogate stochastic model is built using a Bayesian nonparametric approach, Dirichlet process. Probability distributions derived from uncertainty analysis using MELCOR and the stochastic model show good agreement. The appropriateness of the stochastic model is cross-validated through the comparison with MELCOR results. The importance measure of uncertain input variables are calculated according to their influences on the uncertainty distribution as first-order effect and total effect. The validity of the present methodology is demonstrated through an example with three uncertain input variables. - Highlights: • A method of source term uncertainty and sensitivity analysis is proposed. • Source term in Fukushima Daiichi NPP severe accident is demonstrated. • Uncertainty distributions of source terms show non-standard shapes. • A surrogate model for integrated code is constructed by using Dirichlet process. • Importance ranking of influential input variables is obtained

  16. Analysis of the source term formation in a severe accident initiated by end fitting failure in CANDU type reactors

    Constantin, Marin; Constantin, Alina; Apostol, Minodora [Institute for Nuclear Research, Pitesti (Romania)

    2014-04-15

    CANDU type reactors have some peculiarities in initiation and progression of severe accidents. In the present paper one of the specificities - the End Fitting Failure accident - is analysed from the point of view of source term formation. The accident is initiated by a failure of the re-fueling machine. Fuel bundles are ejected in the re-fuelling machine room and fuel elements suffer a significant fragmentation by mechanical impact and by the rapid increase of the temperature. A direct transfer of the fission products occurs directly to the containment. The source term in the containment and also the source term to the environment is calculated supposing an open venting communication to the external atmosphere. The simulation is performed by using the ASTEC code in coupled calculation CPA-IODE-ISODOPE-DOSE option. The evolution of the distributions for the most important released fission products is presented for different regions and for different hosts. The most important factors of influence on the source term formation are identified and discussed. (orig.)

  17. Accidents with orphan sources

    The International Atomic Energy Agency has specifically defined statutory functions relating to the development of standards of safety and the provision for their application. It also has responsibilities placed on it by virtue of a number of Conventions, two of which are relevant to nuclear accidents or radiological emergencies - the Convention on Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency. An overview of the way in which these functions are being applied to prevent and respond to radiological accidents, particularly those involving orphan sources, is described in this paper. Summaries of a number of such accidents and of the Agency's Action Plan relating to the safety and security of radiation sources are given. (orig.)

  18. Water simulation experiments on the instantaneous source term of a severe breeder reactor accident

    FAUST is an experimental program to give contributions to the assessment of the instantaneous source term in case of an LMFBR loss-of-flow accident with expanding fuel or sodium vapor. In the FAUST 1a-series, experiments with discharge of a gas-particle mixture (nitrogen from 0.3 to 2.0 MPa with iron or nickel powder of different particle size) from a 1.45 liter source into a water pool cylinder of 28.8 cm diameter and 1 m height by rupture disks were performed at different pool height (0.90 cm). The system was closed, i.e. no openings were provided in the cover plate. Important measuring instruments were high-speed cameras, pressure transducers and magnets for article trapping in the cover gas. The most important quantity to be determined was the retention factor RF, defined as the ratio of the amount of particles discharged to the amount trapped in the cover gas. Furthermore, the expansion characteristics of the bubble, the correlated cover gas phenomena, the oscillation period and the entrainment were considered. In most cases, particle release stayed below detection limit, which corresponds to RF > 104. For the 1B series, using the same source, a larger pool vessel (63 cm diameter, 60 cm height) was installed and a cover plate with two openings of 4 cm diameter to simulate leaks. The discharge pressure was varied from 0.002 to 4 MPa. Other experimental parameters were pool height (0.50 cm), particles size (1 to 100 μm), and leak size. A release of airborne particles was found only at very low discharge pressure. At high pressure, major amounts of water were released, whereas the release of particles remained below detection limit (retention factor > 104). The oscillation period was of the order of 80 msec for 1A and 50 msec for 1B. Approximative calculations have shown that the large particle absorption may be explained by impaction during the bubble oscillations. (orig.)

  19. Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses; Volume 1, Revision 1

    Gorham, E.D.; Breeding, R.J.; Brown, T.D.; Harper, F.T. [Sandia National Labs., Albuquerque, NM (United States); Helton, J.C. [Arizona State Univ., Tempe, AZ (United States); Murfin, W.B. [Technadyne Engineering Consultants, Inc., Albuquerque, NM (United States); Hora, S.C. [Hawaii Univ., Hilo, HI (United States)

    1993-12-01

    NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community.

  20. Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses. Volume 1, Revision 1

    NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community

  1. Recent advances in the source term area within the SARNET European severe accident research network

    Herranz, L.E., E-mail: luisen.herranz@ciemat.es [Centro de Investigaciones Energeticas Medio Ambientales y Tecnologica, CIEMAT, Avda. Complutense 40, E-28040 Madrid (Spain); Haste, T. [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Kärkelä, T. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT Espoo (Finland)

    2015-07-15

    Highlights: • Main achievements of source term research in SARNET are given. • Emphasis on the radiologically important iodine and ruthenium fission products. • Conclusions on FP release, transport in the RCS and containment behaviour. • Significance of large-scale integral experiments to validate the analyses used. • A thorough list of the most recent references on source term research results. - Abstract: Source Term has been one of the main research areas addressed within the SARNET network during the 7th EC Framework Programme of EURATOM. The entire source term domain was split into three major areas: oxidising impact on source term, iodine chemistry in the reactor coolant system and containment and data and code assessment. The present paper synthesises the main technical outcome stemming from the SARNET FWP7 project in the area of source term and includes an extensive list of references in which deeper insights on specific issues may be found. Besides, based on the analysis of the current state of the art, an outlook of future source term research is outlined, where major changes in research environment are discussed (i.e., the end of the Phébus FP project; the end of the SARNET projects; and the launch of HORIZON 2020). Most probably research projects will be streamlined towards: release and transport under oxidising conditions, containment chemistry, existing and innovative filtered venting systems and others. These will be in addition to a number of projects that have been completed or are ongoing under different national and international frameworks, like VERDON, CHIP and EPICUR started under the International Source Term Programme (ISTP), the OECD/CSNI programmes BIP, BIP2, STEM, THAI and THAI2, and the French national programme MIRE. The experimental PASSAM project under the 7th EC Framework programme, focused on source term mitigation systems, is highlighted as a good example of a project addressing potential enhancement of safety systems

  2. Recent advances in the source term area within the SARNET European severe accident research network

    Highlights: • Main achievements of source term research in SARNET are given. • Emphasis on the radiologically important iodine and ruthenium fission products. • Conclusions on FP release, transport in the RCS and containment behaviour. • Significance of large-scale integral experiments to validate the analyses used. • A thorough list of the most recent references on source term research results. - Abstract: Source Term has been one of the main research areas addressed within the SARNET network during the 7th EC Framework Programme of EURATOM. The entire source term domain was split into three major areas: oxidising impact on source term, iodine chemistry in the reactor coolant system and containment and data and code assessment. The present paper synthesises the main technical outcome stemming from the SARNET FWP7 project in the area of source term and includes an extensive list of references in which deeper insights on specific issues may be found. Besides, based on the analysis of the current state of the art, an outlook of future source term research is outlined, where major changes in research environment are discussed (i.e., the end of the Phébus FP project; the end of the SARNET projects; and the launch of HORIZON 2020). Most probably research projects will be streamlined towards: release and transport under oxidising conditions, containment chemistry, existing and innovative filtered venting systems and others. These will be in addition to a number of projects that have been completed or are ongoing under different national and international frameworks, like VERDON, CHIP and EPICUR started under the International Source Term Programme (ISTP), the OECD/CSNI programmes BIP, BIP2, STEM, THAI and THAI2, and the French national programme MIRE. The experimental PASSAM project under the 7th EC Framework programme, focused on source term mitigation systems, is highlighted as a good example of a project addressing potential enhancement of safety systems

  3. Data assimilation and source term estimation during the early phase of a nuclear accident

    Golubenkov, A.; Borodin, R. [SPA Typhoon, Emergency Centre (Russian Federation); Sohier, A.; Rojas Palma, C. [Centre de l`Etude de l`Energie Nucleaire, Mol (Belgium)

    1996-02-01

    The mathematical/physical base of possible methods to model the source term during an accidental release of radionuclides is discussed. Knowledge of the source term is important in view of optimizing urgent countermeasures to the population. In most cases however, it will be impossible to assess directly the release dynamics. Therefore methods are under development in which the source term is modelled, based on the comparison of off-site monitoring data and model predictions using an atmospheric dispersion model. The degree of agreement between the measured and calculated characteristics of the radioactive contamination of the air and the ground surface is an important criterion in this process. Due to the inherent complexity, some geometrical transformations taking space-time discrepancies between observed and modelled contamination fields are defined before the source term is adapted. This work describes the developed algorithms which are also tested against data from some tracer experiments performed in the past. This method is also used to reconstruct the dynamics of the Chernobyl source term. Finally this report presents a concept of software to reconstruct a multi-isotopic source term in real-time.

  4. Assessment Of Source Term And Radiological Consequences For Design Basis Accident And Beyond Design Basis Accident Of The Dalat Nuclear Research Reactor

    The paper presents results of the assessment of source terms and radiological consequences for the Design Basis Accident (DBA) and Beyond Design Basis Accident (BDBA) of the Dalat Nuclear Research Reactor. The dropping of one fuel assembly during fuel handling operation leading to the failure of fuel cladding and the release of fission products into the environment was selected as a DBA for the analysis. For the BDBA, the introduction of a step positive reactivity due to the falling of a heavy block from the rotating bridge crane in the reactor hall onto a part of the platform where are disposed the control rod drives is postulated. The result of the radiological consequence analyses shows that doses to members of the public are below annual dose limit for both DBA and BDBA events. However, doses from exposure to operating staff and experimenters working inside the reactor hall are predicted to be very high in case of BDBA and therefore the protective actions should be taken when the accident occurs. (author)

  5. Quantification of in-containment fission products source term for 1000 MWe PWR under loss of coolant accident

    Highlights: • Kinetic modeling for in-containment fission product activity. • Modeling and simulation of in-containment source term after LOCA. • Quantification of airborne in-containment activity. • BURNUP activity calculation and comparison with literature. • Study the effect of ESFs and coolant retention with mixing rate. - Abstract: The aim of this work is the modeling and simulation of in-containment fission products (FPs) quantification and behavior under loss of coolant accident (LOCA) in terms of NUREG-1465 key aspects. For this purpose, a kinetic model has been developed to determine the quantification and behavior of in-containment source term after loss of coolant accident for typical 1000 MWe PWR. A more realistic approach of continuous release of fission products from damaged core has been implemented with coolant retention. The simulation for in-containment fission product quantification influenced by containment atmosphere and containment system response has been carried out. Dramatic results have been obtained upon comparison study of fission product behaviors with different computational values. Moreover a contradiction in mixing rate (wx) value has been observed with a factor of 10 in comparison with Saeed et al. (2012)

  6. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  7. A risk-based evaluation of the impact of key uncertainties on the prediction of severe accident source terms - STU

    The purpose of this project is to address the key uncertainties associated with a number of fission product release and transport phenomena in a wider context and to assess their relevance to key severe accident sequences. This project is a wide-based analysis involving eight reactor designs that are representative of the reactors currently operating in the European Union (EU). In total, 20 accident sequences covering a wide range of conditions have been chosen to provide the basis for sensitivity studies. The appraisal is achieved through a systematic risk-based framework developed within this project. Specifically, this is a quantitative interpretation of the sensitivity calculations on the basis of 'significance indicators', applied above defined threshold values. These threshold values represent a good surrogate for 'large release', which is defined in a number of EU countries. In addition, the results are placed in the context of in-containment source term limits, for advanced light water reactor designs, as defined by international guidelines. Overall, despite the phenomenological uncertainties, the predicted source terms (both into the containment, and subsequently, into the environment) do not display a high degree of sensitivity to the individual fission product issues addressed in this project. This is due, mainly, to the substantial capacity for the attenuation of airborne fission products by the designed safety provisions and the natural fission product retention mechanisms within the containment

  8. Krsko source term analysis

    The Krsko Source Term Analysis (STA) has been provided as integral part of the Krsko Individual Plant Examination Project (IPE) Level 2 (Containment Analysis). Based on its own definition, the STA quantifies the magnitude, time dependence and composition of the fission product releases which characterize each Release Category (RC). The Krsko STA also addresses the definition of each RC, identification and the choice of dominant accident sequences within a release category, analysis of the representative accident sequences using MAAP 3.OB (Modular Accident Analysis Program) revision 18 to estimate the source term characteristic and discussion of identified major areas of uncertainty. (author)

  9. Simplified approach for reconstructing the atmospheric source term for Fukushima Daiichi nuclear power plant accident using scanty meteorological data

    Highlights: • Estimation of source terms for I-131 and Cs-137 for Fukushima Daiichi NPP. • Simplified Gaussian puff based atmospheric dispersion model is used. • Good agreement of estimated values as compared to that given by NISA, TEPCO and IRSN. - Abstract: The atmospheric source term for the Fukushima Daiichi nuclear power plant accident in March 2011 has been estimated by a Gaussian puff based atmospheric dispersion model. The scanty meteorological data available at irregular time intervals are utilized to demonstrate the utility of such data along with a simplified modeling approach to derive useful information. The source terms for I-131 and Cs-137 have been estimated as a function of time from the observed values of activity concentration in the air and deposited activity on the ground. The model results suggest that during 12th March 2011–16th March 2011, 9.29 × 1016 Bq of I-131 and 6.15 × 1015 Bq of Cs-137 might have got released to the environment

  10. Source term and radiological consequence evaluation for nuclear accidents using a 'hand type' methodology

    In the last decades, hand type calculations have been replaced by computerized solutions, which are much more accurate, but the preparation of an input to run the code can be a time consuming process and can require a laborious work. This is why, a place for hand calculation based on nomograms still exist in some areas. An example is emergency response to an accidental release of radioactive contaminants when the health of persons close to the accident site might be at risk. In this case, results from computerized accident consequences assessment models may be delayed due to the equipment malfunction or the time required developing minimal input files and performing the calculations (typically more than five minutes). A simple nomogram (developed using computerized dispersion model calculations) can provide dispersion and dose estimates within a minute. The paper presents the methodology used for these 'hand type' calculation and the nomograms, figures and tables used to evaluate the dose to an individual close to the release point. In order to illustrate the use of methodology, a hypothetical severe accident scenario involving 14-MW INR-TRIGA research reactor was considered. (authors)

  11. Source term assessment, containment atmosphere control systems, and accident consequences. Report to CSNI by an OECD/NEA Group of experts

    CSNI Report 135 summarizes the results of the work performed by CSNI's Principal Working Group No. 4 on the Source Term and Environmental Consequences (PWG4) during the period extending from 1983 to 1986. This document contains the latest information on some important topics relating to source terms, accident consequence assessment, and containment atmospheric control systems. It consists of five parts: (1) a Foreword and Executive Summary prepared by PWG4's Chairman; (2) a Report on the Technical Status of the Source Term; (3) a Report on the Technical Status of Filtration and Containment Atmosphere Control Systems for Nuclear Reactors in the Event of a Severe Accident; (4) a Report on the Technical Status of Reactor Accident Consequence Assessment; (5) a list of members of PWG4

  12. Determination of the in-containment source term for a Large-Break Loss of Coolant Accident

    This is the report of a project that focused on one of the most important design basis accidents: the Large Break Loss Of Coolant Accident (LBLOCA) (for pressurised water reactors). The first step in the calculation of the radiological consequences of this accident is the determination of the source term inside the containment. This work deals with this part of the calculation of the LBLOCA radiological consequences for which a previous benchmark (1988) has shown wide variations in the licensing practices adopted by European countries. The calculation of this source term may naturally be split in several steps (see chapter II), corresponding to several physical stages in the release of fission products: fraction of core failure, release from the damaged fuel, airborne part of the release and the release into the reactor coolant system and the sumps, chemical behaviour of iodine in the aqueous and gas phases, natural and spray removal in the containment atmosphere. A chapter is devoted to each of these topics. In addition, two other chapters deal with the basic assumptions to define the accidental sequence and the nuclides to be considered when computing doses associated with the LBLOCA. The report describes where there is agreement between the partner organisations and where there are still differences in approach. For example, there is agreement concerning the percentage of failed fuel which could be used in future licensing assessments (however this subject is still under discussion in France, a lower value is thinkable). For existing plants, AVN (Belgium) wishes to keep the initial licensing assumptions. For the release from damaged fuel, there is not complete agreement: AVN (Belgium) wishes to maintain its present approach. IPSN (France), GRS (Germany) and NNC (UK) prefer to use their own methodologies that result in slightly different values to the proposed values for a common position. There are presently no recommendations of the release of fuel particulates

  13. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104). [PWR; BWR

    Kress, T. S. [comp.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time.

  14. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104)

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time

  15. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.; Folga, S.; Policastro, A.; Freeman, W.; Jackson, R.; Mishima, J.; Turner, S.

    1996-12-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report.

  16. Passive ALWR source term

    The purpose of this report is to provide technical support for the physically-based source term which is proposed as the licensing design basis fission product release from a major core accident for the Passive Advanced Light Water Reactor (ALWR) in Volume 3, Section 5 of the ALWR Requirements Document. A substantial body of new research motivated by the Three Mile Island (TMI) accident is maturing, and the ALWR Requirements Document provides an opportunity to incorporate this experience in an updated source term. This update will provide a more rational basis for Passive ALWR accident mitigation system designs, particularly where the designs afford opportunities for improvement and innovation. Great attention has been paid to accident prevention in the ALWR Requirements Document which will reduce the likelihood of core damage by an order of magnitude or more compared to earlier LWR designs. Nonetheless, for defense-in-depth the Passive ALWR source term is based on evaluation of a core damage event. Selection of this core damage event and the associated quantification of the fission product release were done in a conservative, yet physically-based manner so as to provide significant margin to the expected releases, given an ALWR accident, while avoiding non-physical assumptions which could produce mitigation system designs not well-suited to the important accidents. The physically-based source term presented in this report is intended for use in ALWR design basis analysis defining the radiological environment for plant systems and equipment and evaluating the offsite dose for emergency planning considerations. 100 refs., 18 figs., 44 tabs

  17. An inverse modeling method to assess the source term of the Fukushima nuclear power plant accident using gamma dose rate observations

    O. Saunier

    2013-06-01

    Full Text Available The Chernobyl nuclear accident and more recently the Fukushima accident highlighted that the largest source of error on consequences assessment is the source term including the time evolution of the release rate and its distribution between radioisotopes. Inverse modeling methods, which combine environmental measurements and atmospheric dispersion models, have proven efficient in assessing source term due to an accidental situation (Gudiksen, 1989; Krysta and Bocquet, 2007; Stohl et al., 2012a; Winiarek et al., 2012. Most existing approaches are designed to use air sampling measurements (Winiarek et al., 2012 and some of them also use deposition measurements (Stohl et al., 2012a; Winiarek et al., 2013 but none of them uses dose rate measurements. However, it is the most widespread measurement system, and in the event of a nuclear accident, these data constitute the main source of measurements of the plume and radioactive fallout during releases. This paper proposes a method to use dose rate measurements as part of an inverse modeling approach to assess source terms. The method is proven efficient and reliable when applied to the accident at the Fukushima Daiichi nuclear power plant (FD-NPP. The emissions for the eight main isotopes 133Xe, 134Cs, 136Cs, 137Cs, 137mBa, 131I, 132I and 132Te have been assessed. Accordingly, 103 PBq of 131I, 35.5 PBq of 132I, 15.5 PBq of 137Cs and 12 100 PBq of noble gases were released. The events at FD-NPP (such as venting, explosions, etc. known to have caused atmospheric releases are well identified in the retrieved source term. The estimated source term is validated by comparing simulations of atmospheric dispersion and deposition with environmental observations. The result is that the model-measurement agreement for all of the monitoring locations is correct for 80% of simulated dose rates that are within a factor of 2 of the observed values. Changes in dose rates over time have been overall properly

  18. Review of plutonium aerosol source-term in nuclear accident%核事故条件下钚气溶胶源项研究综述

    刘文杰; 胡八一; 李庆忠

    2011-01-01

    The theoretical and experimental evaluations of plutonium aerosol source-term in nuclear accident are summarized and reviewed in this paper. Hie content of this paper include oxidation mechanism of plutonium, paniculate distribution of plutonium aerosol, aerosol release fraction (ARF) and respirable fraction (RF) of radioactive aerosolization during nuclear accident. The source-term data of three kinds of nuclear accidents which are explosive detonation, static combustion and dynamic combustion have been investigated. The latter two accidental scenes tend to imitate stockpile fire accidents and air transportation disasters wherein the aircrafts and missiles with nuclear devices run up against unexpected fire, crash or in-flight breakup respectively . It is indicated that the aerosolization mechanisms of static combustion and the dynamic combustion without plutonium droplets sparking and explosion are all derived from the oxide particles spilled from the plutonium surface. The size distributions of smaller aerosol particles dispersed with updraft and biggish ones deposited in the soil during static combustion have been measured individually. After that, the full-scale distribution could be obtained in accordance with the combination of the two parts. The investigation of full-scale particu-late distribution in the dynamic combustion scene is unreachable due to the restriction of experimental conditions. The aerosolization distribution in the explosive detonation scene is listed via field test data of Operation Roller Coaster. The source-term of static combustion is lower than the dynamic combustion without sparking and explosion of plutonium droplets, which is based on the thermo dissipation and the loss of oxygen around the plutonium solidity. The dynamic combustion with plutonium droplets sparking and explosion leads to vapor venting and higher source-term. During the explosive detonation the reaction of plutonium oxidation is severest and leads to the highest

  19. Estimation of the time-dependent radioactive source-term from the Fukushima nuclear power plant accident using atmospheric transport modelling

    Schoeppner, M.; Plastino, W.; Budano, A.; De Vincenzi, M.; Ruggieri, F.

    2012-04-01

    Several nuclear reactors at the Fukushima Dai-ichi power plant have been severely damaged from the Tōhoku earthquake and the subsequent tsunami in March 2011. Due to the extremely difficult on-site situation it has been not been possible to directly determine the emissions of radioactive material. However, during the following days and weeks radionuclides of 137-Caesium and 131-Iodine (amongst others) were detected at monitoring stations throughout the world. Atmospheric transport models are able to simulate the worldwide dispersion of particles accordant to location, time and meteorological conditions following the release. The Lagrangian atmospheric transport model Flexpart is used by many authorities and has been proven to make valid predictions in this regard. The Flexpart software has first has been ported to a local cluster computer at the Grid Lab of INFN and Department of Physics of University of Roma Tre (Rome, Italy) and subsequently also to the European Mediterranean Grid (EUMEDGRID). Due to this computing power being available it has been possible to simulate the transport of particles originating from the Fukushima Dai-ichi plant site. Using the time series of the sampled concentration data and the assumption that the Fukushima accident was the only source of these radionuclides, it has been possible to estimate the time-dependent source-term for fourteen days following the accident using the atmospheric transport model. A reasonable agreement has been obtained between the modelling results and the estimated radionuclide release rates from the Fukushima accident.

  20. Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

    Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

  1. Scientists debate nuclear source terms

    The nuclear source term, defined as the quantity, timing, and characteristic of the release of radioactive material to the environment following a core-melt accident, was thoroughly debated in 1985. This debate, summarized here, turns on the Nuclear Regulatory Commission's (NRC) source term for radioactive iodine, which is postulated as potentially the most life-threatening radionuclide that might escape in a nuclear power-plant accident. Following the Three Mile Island (TMI) accident, from which only traces of radioiodine escaped, scientists began arguing that nuclear regulations based on source-term calculations are erroneous and should be modified. The American Nuclear Society (ANS) and industry researchers have concluded that warranted reductions in the NRC source terms could range from a factor of ten to several factors of ten in most accident scenarios. The American Physical Society (APS), after agreeing with a large body of the conclusions from the other research groups, has told NRC that its source-term data base is still inadequate because of the existence of a number of uncertainties it found therein. Although APS presented no such conclusion, its findings made clear to NRC that an early reduction of all source terms is not warranted. The anti-nuclear lobby agrees with APS. The NRC has taken a cautious, conservative approach to the revision of its regulations based on new source-term data, although it too concedes that its old methodologies and conclusions must be revised and ultimately superseded

  2. 77 FR 19740 - Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident

    2012-04-02

    ...-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray... accident. II. Further Information DG-1234 was published in the Federal Register on July 15 2010, (75...

  3. Supplemental analysis of accident sequences and source terms for waste treatment and storage operations and related facilities for the US Department of Energy waste management programmatic environmental impact statement

    This report presents supplemental information for the document Analysis of Accident Sequences and Source Terms at Waste Treatment, Storage, and Disposal Facilities for Waste Generated by US Department of Energy Waste Management Operations. Additional technical support information is supplied concerning treatment of transuranic waste by incineration and considering the Alternative Organic Treatment option for low-level mixed waste. The latest respirable airborne release fraction values published by the US Department of Energy for use in accident analysis have been used and are included as Appendix D, where respirable airborne release fraction is defined as the fraction of material exposed to accident stresses that could become airborne as a result of the accident. A set of dominant waste treatment processes and accident scenarios was selected for a screening-process analysis. A subset of results (release source terms) from this analysis is presented

  4. On the potential limitation of radiological source term releases considering severe core accidents in future PWR plants

    Future nuclear power plants should be so safe that even in case of such a severe accident there will be no need of drastic external disaster control measures such as an evacuation or resettlement of the population from the vicinity of a nuclear power plant. It is shown by the example of a future 1400 MWe pressurized water reactor plant that this goal can be attained in principle by providing a double containment with the annulus vented via an appropriate emergency standby filter. Within the framework of disaster prevention a set of parameters for accident conditions is elaborated under which the lower levels of intervention for evacuation are not attained. (orig./HP)

  5. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    Japan Nuclear Energy Safety organization (JNES) is developing severe accident analysis codes in order to apply to the probabilistic safety assessment (PSA) for a typical fast breeder reactor (FBR). The AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary and the release fraction to the environment of fission products (FP). This report summarized results analyzed using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass (CVBP) scenario, and the containment failure scenario due to hydrogen deflagration or detonation. The results showed that the coolant temperature of the primary system and the secondary system in the PLOHS sequence increased at the almost same temperature, and the creep damage to the reactor coolant boundary became significant when coolant temperature exceeded about 1,100 K. The release fractions of FP in the CVBP case were estimated to be 0.99 for Xe, 0.14 for iodine, 0.44 for Cs and 0.01 for non-volatile tetravalent Ce. The release fractions of FP in the containment vessel failure case due to hydrogen burning were estimated to be 0.82 for Xe, 0.06 for iodine, 0.06 for Cs and 0.003 for non-volatile tetravalent Ce. In the present study, release fractions of FPs to the environment were obtained for the CVBP and the containment failure cases of the PLOHS accident sequence for the typical FBR plant. (author)

  6. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  7. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J. [and others

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report.

  8. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  9. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    JNES is developing severe accident analysis codes in order to apply to the probability safety analysis (PSA) for a typical fast breeder reactor (FBR). AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary, and the discharge rate to the environment of fission products (FP). This report summarizes analysis results using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass scenario (CVBP) and the containment failure scenario by hydrogen deflagration or detonation. The coolant temperature of the primary system and the secondary system in the PLOHS sequence increases at the almost same temperature, and the creep damage to the reactor coolant boundary will become remarkable if coolant temperature exceeds about 1,100 K. In the CVBP scenario, when an intermediate heat exchanger is ruptured by creep and the boundary of the secondary system is failed, the path from the primary system to environment is formed. Then, the reactor vessel (RV) is failed and sodium in the primary coolant system releases into the reactor vessel room (RV room). Sodium of high temperature which fell in the RV room damages the floor liner, and generates hydrogen by a reaction with concrete. In addition the reactor core is exposed into atmosphere and the core temperature increases with decay heat and then volatile FP and non-volatile FP are released to the environment through the secondary system from the primary system. In the non-CVBP scenario which the intermediate heat exchanger does not fail by creep, core debris falls into the RV room after reactor vessel failure or evaporation of sodium coolant molten. FPs released from the reactor vessel are retained in the RV room, the primary system room, the containment dome and so on. The hydrogen generated by sodium-concrete reaction and

  10. Source term estimation using air concentration measurements and a Lagrangian dispersion model - Experiments with pseudo and real cesium-137 observations from the Fukushima nuclear accident

    Chai, Tianfeng; Draxler, Roland; Stein, Ariel

    2015-04-01

    A transfer coefficient matrix (TCM) was created in a previous study using a Lagrangian dispersion model to provide plume predictions under different emission scenarios. The TCM estimates the contribution of each emission period to all sampling locations and can be used to estimate source terms by adjusting emission rates to match the model prediction with the measurements. In this paper, the TCM is used to formulate a cost functional that measures the differences between the model predictions and the actual air concentration measurements. The cost functional also includes a background term which adds the differences between a first guess and the updated emission estimates. Uncertainties of the measurements, as well as those for the first guess of source terms are both considered in the cost functional. In addition, a penalty term is added to create a smooth temporal change in the release rate. The method is first tested with pseudo observations generated using the Hybrid Single Particle Lagrangian Integrated Trajectory (HYSPLIT) model at the same location and time as the actual observations. The inverse estimation system is able to accurately recover the release rates and performs better than a direct solution using singular value decomposition (SVD). It is found that computing ln(c) differences between model and observations is better than using the original concentration c differences in the cost functional. The inverse estimation results are not sensitive to artificially introduced observational errors or different first guesses. To further test the method, daily average cesium-137 air concentration measurements around the globe from the Fukushima nuclear accident are used to estimate the release of the radionuclide. Compared with the latest estimates by Katata et al. (2014), the recovered release rates successfully capture the main temporal variations. When using subsets of the measured data, the inverse estimation method still manages to identify most of the

  11. Atmospheric discharge and dispersion of radionuclides during the Fukushima Dai-ichi Nuclear Power Plant accident, 2; Verification of the source term and analysis of regional-scale atmospheric dispersion

    寺田 宏明; 堅田 元喜; 茅野 政道; 永井 晴康

    2012-01-01

    Regional-scale atmospheric dispersion simulations were carried out to verify source term of 131I and 137Cs estimated by our previous studies and analyze the atmospheric dispersion during the Fukushima Dai-ichi Nuclear Power Plant accident with measurements of daily and monthly surface depositions over land in Eastern Japan from March 12 to April 30, 2011. The prediction accuracy of daily surface deposition by using the refined source term was mostly within a factor of 10 without apparent bias...

  12. Detailed source term estimation of atmospheric release during the Fukushima Dai-ichi nuclear power plant accident by coupling atmospheric and oceanic dispersion models

    Katata, Genki; Chino, Masamichi; Terada, Hiroaki; Kobayashi, Takuya; Ota, Masakazu; Nagai, Haruyasu; Kajino, Mizuo

    2014-05-01

    Temporal variations of release amounts of radionuclides during the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) accident and their dispersion process are essential to evaluate the environmental impacts and resultant radiological doses to the public. Here, we estimated a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data and coupling atmospheric and oceanic dispersion simulations by WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN developed by the authors. New schemes for wet, dry, and fog depositions of radioactive iodine gas (I2 and CH3I) and other particles (I-131, Te-132, Cs-137, and Cs-134) were incorporated into WSPEEDI-II. The deposition calculated by WSPEEDI-II was used as input data of ocean dispersion calculations by SEA-GEARN. The reverse estimation method based on the simulation by both models assuming unit release rate (1 Bq h-1) was adopted to estimate the source term at the FNPP1 using air dose rate, and air sea surface concentrations. The results suggested that the major release of radionuclides from the FNPP1 occurred in the following periods during March 2011: afternoon on the 12th when the venting and hydrogen explosion occurred at Unit 1, morning on the 13th after the venting event at Unit 3, midnight on the 14th when several openings of SRV (steam relief valve) were conducted at Unit 2, morning and night on the 15th, and morning on the 16th. The modified WSPEEDI-II using the newly estimated source term well reproduced local and regional patterns of air dose rate and surface deposition of I-131 and Cs-137 obtained by airborne observations. Our dispersion simulations also revealed that the highest radioactive contamination areas around FNPP1 were created from 15th to 16th March by complicated interactions among rainfall (wet deposition), plume movements, and phase properties (gas or particle) of I-131 and release rates

  13. Analysis of accident sequences and source terms at treatment and storage facilities for waste generated by US Department of Energy waste management operations. Volume 1: Sections 1-9

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the US Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies assessed, and the resultant radiological and chemical source terms evaluated. A personal-computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for the calculation of human health risk impacts. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated, and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. Key assumptions in the development of the source terms are identified. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also discuss specific accident analysis data and guidance used or consulted in this report

  14. Atmospheric discharge and dispersion of radionuclides during the Fukushima Dai-ichi Nuclear Power Plant accident. Part I: Source term estimation and local-scale atmospheric dispersion in early phase of the accident

    The atmospheric release of 131I and 137Cs in the early phase of the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) accident from March 12 to 14, 2011 was estimated by combining environmental data with atmospheric dispersion simulations under the assumption of a unit release rate (1 Bq h−1). For the simulation, WSPEEDI-II computer-based nuclear emergency response system was used. Major releases of 131I (>1015 Bq h−1) were estimated when air dose rates increased in FNPP1 during the afternoon on March 12 after the hydrogen explosion of Unit 1 and late at night on March 14. The high-concentration plumes discharged during these periods flowed to the northwest and south–southwest directions of FNPP1, respectively. These plumes caused a large amount of dry deposition on the ground surface along their routes. Overall, the spatial pattern of 137Cs and the increases in the air dose rates observed at the monitoring posts around FNPP1 were reproduced by WSPEEDI-II using estimated release rates. The simulation indicated that air dose rates significantly increased in the south–southwest region of FNPP1 by dry deposition of the high-concentration plume discharged from the night of March 14 to the morning of March 15. - Highlights: ► Source term during the Fukushima Dai-ichi Nuclear Power Plant accident was estimated. ► Atmospheric dispersion simulation was carried out for estimation. ► Major releases were estimated in the afternoon on March 12 and the night on March 14. ► Air dose rate increased due to dry deposition during the night of March 14.

  15. Source term uncertainties for the Seabrook station probabilistic risk assessment

    Accident source terms are defined as a set of release categories that define the time and quantity of radionuclides released for different containment failure modes and groups of accident sequences. Best estimates of these source terms are characterized by very large uncertainties. A probabilistic risk assessment (PRA) aims to determine realistic consequences. Estimating source term uncertainties is therefore as important as estimating the source terms themselves. A systematic analysis of source term uncertainties was performed for the Seabrook PRA. Uncertainties were quantified due to in-vessel retention, ex-vessel retention, different accident sequences, accident progression phenomena, containment failure time, containment failure modes, and leak path attenuation

  16. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 1: Sections 1-9

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  17. Atmospheric discharge and dispersion of radionuclides during the Fukushima Dai-ichi Nuclear Power Plant accident. Part II: verification of the source term and analysis of regional-scale atmospheric dispersion

    Regional-scale atmospheric dispersion simulations were carried out to verify the source term of 131I and 137Cs estimated in our previous studies, and to analyze the atmospheric dispersion and surface deposition during the Fukushima Dai-ichi Nuclear Power Plant accident. The accuracy of the source term was evaluated by comparing the simulation results with measurements of daily and monthly surface depositions (fallout) over land in eastern Japan from March 12 to April 30, 2011. The source term was refined using observed air concentrations of radionuclides for periods when there were significant discrepancies between the calculated and measured daily surface deposition, and when environmental monitoring data, which had not been used in our previous studies, were now available. The daily surface deposition using the refined source term was predicted mostly to within a factor of 10, and without any apparent bias. Considering the errors in the model prediction, the estimated source term is reasonably accurate during the period when the plume flowed over land in Japan. The analysis of regional-scale atmospheric dispersion and deposition suggests that the present distribution of a large amount of 137Cs deposition in eastern Japan was produced primarily by four events that occurred on March 12, 15–16, 20, and 21–23. The ratio of wet deposition to the total varied widely depending on the influence by the particular event. - Highlights: ► Source term during the Fukushima Dai-ichi Nuclear Power Plant accident was verified. ► Accuracy of surface deposition by atmospheric dispersion simulations was evaluated. ► The source term of 131I and 137Cs was refined to explain the deposition measurements. ► The source term was reasonably accurate while plumes flowed over land in Japan. ► Primary 137Cs deposition over land occurred on March 12, 15–16, 20, and 21–23.

  18. Some practical implications of source term reassessment

    This report provides a brief summary of the current knowledge of severe accident source terms and suggests how this knowledge might be applied to a number of specific aspects of reactor safety. In preparing the report, consideration has been restricted to source term issues relating to light water reactors (LWRs). Consideration has also generally been restricted to the consequences of hypothetical severe accidents rather than their probability of occurrence, although it is recognized that, in the practical application of source term research, it is necessary to take account of probability as well as consequences. The specific areas identified were as follows: Exploration of the new insights that are available into the management of severe accidents; Investigating the impact of source term research on emergency planning and response; Assessing the possibilities which exist in present reactor designs for preventing or mitigating the consequences of severe accidents and how these might be used effectively; Exploring the need for backfitting and assessing the implications of source term research for future designs; and Improving the quantification of the radiological consequences of hypothetical severe accidents for probabilistic safety assessments (PSAs) and informing the public about the realistic risks associated with nuclear power plants. 7 refs

  19. An uncertainty analysis of the hydrogen source term for a station blackout accident in Sequoyah using MELCOR 1.8.5

    Gauntt, Randall O.; Bixler, Nathan E.; Wagner, Kenneth Charles.

    2014-03-01

    A methodology for using the MELCOR code with the Latin Hypercube Sampling method was developed to estimate uncertainty in various predicted quantities such as hydrogen generation or release of fission products under severe accident conditions. In this case, the emphasis was on estimating the range of hydrogen sources in station blackout conditions in the Sequoyah Ice Condenser plant, taking into account uncertainties in the modeled physics known to affect hydrogen generation. The method uses user-specified likelihood distributions for uncertain model parameters, which may include uncertainties of a stochastic nature, to produce a collection of code calculations, or realizations, characterizing the range of possible outcomes. Forty MELCOR code realizations of Sequoyah were conducted that included 10 uncertain parameters, producing a range of in-vessel hydrogen quantities. The range of total hydrogen produced was approximately 583kg 131kg. Sensitivity analyses revealed expected trends with respected to the parameters of greatest importance, however, considerable scatter in results when plotted against any of the uncertain parameters was observed, with no parameter manifesting dominant effects on hydrogen generation. It is concluded that, with respect to the physics parameters investigated, in order to further reduce predicted hydrogen uncertainty, it would be necessary to reduce all physics parameter uncertainties similarly, bearing in mind that some parameters are inherently uncertain within a range. It is suspected that some residual uncertainty associated with modeling complex, coupled and synergistic phenomena, is an inherent aspect of complex systems and cannot be reduced to point value estimates. The probabilistic analyses such as the one demonstrated in this work are important to properly characterize response of complex systems such as severe accident progression in nuclear power plants.

  20. Comprehensive Review of Source Term Analysis and Experimental Programs

    Khurram Mehboob; Cao Xinrong; Majid Ali

    2012-01-01

    Source term evaluation is important in legislation for the licensing of a plant. So a comprehensive review of historical progress in source term analysis and source term experimental programs have been presented for better understanding of core melt progression and fission product behavior under severe accident. The discussions mainly focus on the advancement in source term and comparison between source term experimental programs. For In-vessel and Ex-vessel source term estimation, simulated ...

  1. A Canadian perspective on the source term

    There is a continuing commitment in Canada to perform research to improve our knowledge of fission-product behaviour, and to understand the consequences of possible nuclear reactor accidents. This work was given additional justification by the accident at the Three Mile Island II nuclear generating station. During this accident, the release of radioactive iodine to the environment was low compared to the predictions of the Reactor Safety Study (WASH-1400), a study known to be overly conservative in some areas. There has been considerable effort, through both experiments and computer modelling, to re-evaluate the technical bases for predicting fission-product releases during nuclear reactor accidents. The research work is conventionally referred to under the general title of 'source-term' research. The source term is the amount and type of radioactive material that escapes the boundary of a nuclear reactor containment building, and thus could result in possible exposure of the general public. The source term depends on the type of reactor and the accident sequence. In general, fission-product release to the environment depends on the the airborne fission product concentrations within the containment building. This, in turn, depends on all the physical and chemical factors controlling the release from fuel, migration and transport, chemical speciation, surface interactions, and aerosol behaviour. The scope of the factors influencing the behaviour of fission products within the containment is sufficiently broad to include virtually all the research relevant to reactor safety. It includes studies of thermalhydraulics, fission-product releases from fuel under transient conditions, aerosol behaviour, fission-product chemistry, gas/solution partitoning and fission-product/surface interactions. This report will summarize four areas: (1) the use of the source term in the licensing of nuclear power reactors; (2) the methods used to determine the source term; (3) problems in

  2. The Source Term Code Package

    The Source Term Code Package (STCP) is a set of computer codes which allows analyses of nuclear reactor accidents to produce predictions of fission product release to the environment as a function of reactor design and specifications for the assumed accident. The codes are basically those used in the analyses performed for the BMI-2104 report but they have been combined, improved and streamlined for easier use. The objective in preparing this code package was to make the calculations more direct, traceable and user-independent, with documentation for release to the public. It is important to note that the STCP is not intended to be a research tool but a code for general use in making source term predictions that has a sound and definable basis and produces reasonably accurate results in comparison with more detailed codes. The STCP has four major elements. The overall thermal-hydraulics is provided by the MARCH-3 code which combines the previously separate codes MARCH2, CORCON Mod 2 and CORSORM. Release of fission products and aerosols during core/concrete interactions is predicted with the VANESA code. Detailed thermal-hydraulics and fission product transport in the reactor coolant system are provided by the TRAPMELT3 code formed by combining the previously separate TRAPMELT and MERGE codes. Finally, fission product transport in the containment is predicted by the NAUA-4 code as modified to include fission product removal by pressure suppression pools (SPARC code) and within ice compartments (ICEDF code). The fission product and aerosol groups being tracked in the code package are noble gases, iodine, caesium, tellurium, barium, strontium, ruthenium, lanthanum and cerium groups. Additionally, in-vessel-produced aerosols and ex-vessel-produced aerosols are tracked. The code package produces time-dependent locational distributions, physical forms and transport rates for these groups throughout the course of the accident. (author)

  3. Source terms in relation to air cleaning

    There are two sets of source terms for consideration in air cleaning, those for routine releases and those for accident releases. With about 1000 reactor years of commercial operating experience in the US done, there is an excellent data base for routine and expected transient releases. Specifications for air cleaning can be based on this body of experience with confidence. Specifications for air cleaning in accident situations is another matter. Recent investigations of severe accident behavior are offering a new basis for source terms and air cleaning specifications. Reports by many experts in the field describe an accident environment notably different from previous models. It is an atmosphere heavy with aerosols, both radioactive and inert. Temperatures are sometimes very high; radioiodine is typically in the form of cesium iodide aerosol particles; other nuclides, such as tellurium, are also important aerosols. Some of the present air cleaning requirements may be very important in light of these new accident behavior models. Others may be wasteful or even counterproductive. The use of the new data on accident behavior models to reevaluate requirements promptly is discussed

  4. Source term uncertainties for the Seabrook station probabilistic risk assessment

    Accident source terms are defined as a set of release categories that define the time and quantity of radionuclides released for different containment failure modes and groups of accident sequences. Best estimates of these source terms are characterized by very large uncertainties. A probabilistic risk assessment (PRA) aims to determine realistic consequences. Estimating source term uncertainties is therefore as important as estimating the source terms themselves. A systematic analysis of source term uncertainties was performed for the Seabrook PRA. Uncertainties were quantified due to in-vessel retention, ex-vessel retention, different accident sequences, accident progression phenomena, containment failure time, containment failure modes, and leak path attenuation. The cumulative effect of all sources of uncertainties resulted in a difference between best estimate release fractions and the 99% confidence level of a factor of 10 to 100. The CORRAL predicted release fractions are found to be upper bounds with a 99% confidence level or higher

  5. Source term uncertainties for the seabrook station probabilistic risk assessment

    Accident source terms are defined as a set of release categories that define the time and quantity of radionuclides released for different containment failure modes and groups of accident sequences. Best estimates of these source terms are characterized by very large uncertainties. A probabilistic risk assessment (PRA) aims to determine realistic consequences. Estimating source term uncertainties is therefore as important as estimating the source terms themselves. A systematic analysis of source term uncertainties was performed for the Seabrook PRA. Uncertainties were quantified due to in-vessel retention, ex-vessel retention, different accident sequences, accident progression phenomena, containment failure time, containment failure modes, and leak path attenuation. The cumulative effect of all sources of uncertainties resulted in a difference between best estimate release fractions and the 99% confidence level of a factor of 10 to 100. The CORRAL predicted release fractions are found to be upper bounds with a 99% confidence level or higher

  6. Calculation of source terms for NUREG-1150

    The source terms estimated for NUREG-1150 are generally based on the Source Term Code Package (STCP), but the actual source term calculations used in computing risk are performed by much smaller codes which are specific to each plant. This was done because the method of estimating the uncertainty in risk for NUREG-1150 requires hundreds of source term calculations for each accident sequence. This is clearly impossible with a large, detailed code like the STCP. The small plant-specific codes are based on simple algorithms and utilize adjustable parameters. The values of the parameters appearing in these codes are derived from the available STCP results. To determine the uncertainty in the estimation of the source terms, these parameters were varied as specified by an expert review group. This method was used to account for the uncertainties in the STCP results and the uncertainties in phenomena not considered by the STCP

  7. Comprehensive Review of Source Term Analysis and Experimental Programs

    Khurram Mehboob

    2012-08-01

    Full Text Available Source term evaluation is important in legislation for the licensing of a plant. So a comprehensive review of historical progress in source term analysis and source term experimental programs have been presented for better understanding of core melt progression and fission product behavior under severe accident. The discussions mainly focus on the advancement in source term and comparison between source term experimental programs. For In-vessel and Ex-vessel source term estimation, simulated and design based experiments had been performed. Most of the experiments have been performed to understand the behavior and analysis of radionuclides during different accident consequences. Data obtained from these tests have been used for validation of Computer based codes. This study gives a comprehensive review of the source term and the source term experimental programs designed for investing the source term and fission product release behavior.

  8. Design applications for proposed new source terms

    In October, 1985, the revised Source Terms Application Task Force was formed at Duke Power Company to review potential applications for the numerous government and industry source term research programs. These programs all indicate that radioisotopic releases from severe light water reactor power plant accidents are chemically different and mostly lower than previously predicted. The near term potential applications of this information include reducing or grading Emergency Planning actions and improving public perception of nuclear power plant safety, eliminating or reducing in-plant radiation environments for equipment qualification and post-accident personnel access and redesign of systems and hardware in the plant to more effectively reduce expected radioisotopic releases. The purpose of this paper is to review the preliminary analysis performed for Duke Power's Catawba Nuclear Station to determine potential impact of new source term methods on gamma radiation equipment qualification inside containment and in air filtration systems. The analysis includes: (1) selection of PRA sequences for the specific source term application; (2) use of the MAAP computer code to distribute sources inside containment; (3) new approaches for modeling containment to apply source term forms; and (4) buildup of radioisotopes on air filtration system filters

  9. Definition of loss-of-coolant accident radiation source

    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 deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist

  10. ITER task title - source term data, modelling, and analysis. ITER subtask no. S81TT05/5 (SEP 1-1). Global tritium source term analysis basis document. Subtask 2: tritium releases due to accident conditions. Final task report

    This report presents the methodology and the key assumptions that are adopted in preparing the preliminary estimates of accidental tritium release terms. A room-by room map/table, at the subsystem level, identifying the major equipment and their failure modes, secondary containments, maximum releasable tritium inventory, duration of tritium releases and the tritium release pathways etc. is also included. The tritium release calculations and the room-by-room map/table have been prepared in EXCEL spreadsheets, so that the estimates can be refined easily, and the approach is relatively adaptable to changes of the ITER design information. (author). 22 refs., 7 tabs

  11. Source term estimation for small sized HTRs

    Accidents which have to be considered are core heat-up, reactivity transients, water of air ingress and primary circuit depressurization. The main effort of this paper belongs to water/air ingress and depressurization, which requires consideration of fission product plateout under normal operation conditions; for the latter it is clearly shown, that absorption (penetration) mechanisms are much less important than assumed sometimes in the past. Source term estimation procedures for core heat-up events are shortly reviewed; reactivity transients are apparently covered by them. Besides a general literature survey including identification of areas with insufficient knowledge this paper contains some estimations on the thermomechanical behaviour of fission products in water in air ingress accidents. Typical source term examples are also presented. In an appendix, evaluations of the AVR experiments VAMPYR-I and -II with respect to plateout and fission product filter efficiency are outlined and used for a validation step of the new plateout code SPATRA. (orig.)

  12. Fission product source term research at Oak Ridge National Laboratory

    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, accident sequence progression, etc., will be discussed

  13. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an atmospheric dispersion model with an improved deposition scheme and oceanic dispersion model

    Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

    2015-01-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric and oceanic transport, dispersion and deposition models. The atmospheric and oceanic models used are WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN-FDM (Finite difference oceanic dispersion model), both developed by the authors. A sophisticated deposition scheme, which deals with dry and fog-water depositions, cloud condensation nuclei (CCN) activation, and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The results revealed that the major releases of radionuclides due to the FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, midnight of 14 March when the SRV (safety relief valve) was opened three times at Unit 2, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates. The simulation by WSPEEDI-II using the new source term reproduced the local and regional patterns of cumulative

  14. Regulatory implications of source term studies

    A possible set of criteria for the reassessment of severe accidents in LWRs is discussed as well as the definition of new upper bound generalized source terms for regulatory purposes, as a consequence of the indications given by recent source term studies. Also described is the outcome of recent work carried out at the Comitato Nazionale per la Ricerca e per lo Sviluppo dell'Energia Nucleare e delle Energie Alternative (ENEA/DISP), Italy, by an internal task force and by external supporting organizations. Probabilistic evaluations and qualitative-good judgement criteria are used to define an optimum accident reference level and the corresponding source terms. The approach suggests that only dominant sequences should be considered, neglecting the most unlikely lesser known phenomena, and that the possibility of pre-existing openings in the containment and of operator recovery and repair actions, besides human errors, should be included. Probabilistic evaluations are, in the first instance, used to discriminate between those sequences to be included and those to be neglected. To this purpose a relative cut-off probability, conditional upon core melt, is suggested. A relative threshold of 0.05 on each sequence is used. The exercise is limited to some typical accident sequences. As a result, external releases of volatile fission products (except noble gases) are conservatively found to be lower than 10-3 times the core inventory for both typical PWRs and BWRs. Only a summary of the ENEA/DISP studies and results is provided. (author)

  15. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

    G. Katata

    2014-06-01

    Full Text Available Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1 accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information, and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging for radioactive iodine gas (I2 and CH3I and other particles (CsI, Cs, and Te, was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal

  16. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

    Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

    2014-06-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information), and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN) activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve) at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of

  17. Long-term follow-up of radiation accident patients in Peru: Review of two cases

    Overexposure to radioactive sources used in radiotherapy or industrial radiography may result in severe health consequences. This report assesses the initial clinical status and the medical and psychological long-term follow-up of two radiation accident patients from Peru during the mid-to-late 1990's: one patient exposed to a radiotherapy 60Co source in Arequipa, the other patient to a 192Ir source in Yanango. Commonalities and differences are described. The main causes in both accidents were human error and the failure to apply appropriate safety guidelines and standard operating procedures. Education and training of the personnel working with radiation sources are essential to prevent accidents. The experience gained from the medical management of the two patients is valuable for future treatment of such patients. (authors)

  18. Risk assessment for long-term post-accident sequences

    Probabilistic risk analysis, currently conducted by the CEA (French Atomic Energy Commission) for the French replicate series of 900 MWe power plants, has identified accident sequences requiring long-term operation of some systems after the initiating event. They have been named long-term sequences. Quantification of probabilities of such sequences cannot rely exclusively on equipment failure-on-demand data: it must also take into account operating failures, the probability of which increase with time. Specific studies have therefore been conducted for a number of plant systems actuated during these long-term sequences. This has required: - Definition of the most realistic equipment utilization strategies based on existing emergency procedures for 900 MWe French plants. - Evaluation of the potential to repair failed equipment, given accessibility, repair time, and specific radiation conditions for the given sequence. - Definition of the event bringing the long-term sequence to an end. - Establishment of an appropriate quantification method, capable of taking into account the evolution of assumptions concerning equipment utilization strategies or repair conditions over time. The accident sequence quantification method based on realistic scenarios has been used in the risk assessment of the initiating event loss of reactor coolant accident occurring at power and at shutdown. Compared with the results obtained from conventional methods, this method redistributes the relative weight of accident sequences and also demonstrates that the long term can be a significant contribution to the probability of core melt

  19. Unconventional sources of plant information for accident management

    Oehlberg, R.; Machiels, A.; Chao, J.; Weiss, J. (Electric Power Research Inst., Palo Alto, CA (United States)); True, D.; James, R. (ERIN Engineering and Research, Walnut Creek, CA (United States))

    1992-01-01

    One phase of accident management covers the actions taken during the course of an accident by the plant operating and technical staff to prevent or minimize off-site radiation releases, gain control, and return the plant to a safe state. Inherent in accomplishing these goals is obtaining a clear picture of the nature of the accident and plant status. Development of a consistent and coherent understanding of the accident and plant status requires plant staff to evaluate and interpret data from a wide range of sources. Plant information during an accident can be obtained from the following sources: (1) plant instrumentation, including Regulatory Guide 1.97 instrumentation; and (2) information sources identified in abnormal operations or emergency operations procedures. Probabilistic risk analyses have shown that events involving the loss of key electrical support systems can be significant contributors to core damage. Such events could jeopardize or degrade instrument availability. Plant-specific accident procedures and interpretation of instruments intended for design-basis events may not be applicable in severe accidents. Information sources such as other nuclear steam supply systems (NSSSs) and balance-of-plant (BOP) instrumentation may be available.

  20. PSA modeling of long-term accident sequences

    In the traditional Level 1 PSA, the long term of the accident sequences is usually taken into account in a simplified manner. For example, some of the mitigations which are needed at long term are taken into account in the PSA, but the analysis and the associated failures probabilities quantification are estimated based on generic assessments. In the context of the extension of PSA scope to include the external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the long term of accident sequences induced by initiators which affect the whole site containing several nuclear installations (reactors, fuel pools, ...). This is an essential prerequisite for the development of external hazards PSA. It has to be noted that in the French PSA, even before Fukushima, this type of accident sequences was already taken into account, many insight being used, as complementary information, to enhance the safety level of the plants. The recent French and international operating experience is an opportunity for tuning the actual PSA methods for long term accident sequences modeling. The paper presents the main results of the ongoing efforts in this area. (author)

  1. Influence of Chemistry on source term assessment

    The major goal of a phenomenology analysis of containment during a severe accident situation can be splitedd into the following ones: to know the containment response to the different loads and to predict accurately the fission product and aerosol behavior. In this report, the main results coming from the study of a hypothetical accident scenario, based on LA-4 experiment of LACE project, are presented. In order to do it, several codes have been coupled: CONTEMPT4/MOD5 (thermalhydraulics), NAUA/MOD5 (aerosol physics) and IODE (iodine chemistry). 12 refs. It has been demonstrated the impossibility of assessing with confidence the Source Term if the chemical conduct of some radionuclides is not taken into account. In particular, the influence on the iodine retention efficiency of the sump of variables such as pH has been proven. (Author). 12 refs

  2. Source term evaluation for accident conditions

    The Symposium presentations were divided into 5 sessions devoted to the following topics: in-vessel-release (12 papers), retention in the primary circuit (9 papers), ex-vessel release (6 papers), retention in containment (8 papers) and release from the plant (9 papers). In addition, a poster session was held (8 papers) as well as two panel discussions on the following subjects: containment challenges and regulatory implications. The Proceedings contain all the introductory summaries, papers and posters that were presented during the Symposium. A separate abstract was prepared for each of these papers

  3. Introduction of new terms and lessons for radiological protection after Fukushima Dai-Ichi accident

    The nuclear accidents in the world are very few among various types of operating facilities. However when an accident happened, we have learnt a lot to improve the philosophy, term, definitions, document preparation, equipment's requirement, supporting systems, awareness program and restriction etc. After Fukushima Dai-ichi we have learnt a lot, in this view this paper has been prepared to discuss for radiological protection aspects. Discussion: The probability of nuclear accidents is negligible but when happens, it opens new doors of lessons for radiological protection practices for occupational workers, emergency workers for damage control to prevent catastrophic situation/rescue to life saving actions and the member of the public. The Chernobyl and Three Mile Island accidents have provided a lot experiences for management of emergency situations, documentation, radiation emergency preparedness, emergency equipment's, concept of defense-in-depth, emergency planning zone (EPZ), accidental dose limits, estimation of source term and public dose, intervention levels, decision supporting system, remedial actions in public domain; decontamination of person, houses/building and land and etc. Recent Fukushima Dai-ichi accident in Japan was managed in appreciable manner but still new definitions and lessons for radiological protection have been emerged out. The present paper discusses difficulties w. r. t. the radiological aspects observed/faced by Japanese during nuclear crises. The accident introduced new terms as Natural Dose Rate Unit (NDRU), voluntary evacuation, deliberate evacuation area, restricted area and difference between evacuation zone and EPZ. The Fukushima accident has enforced worldwide regulators and operators to review the individual dose limit and amendment for raise in the dose limit during accident, availability of efficient/adequate quantities of personal dosimeter in public domain, collection arrangement of bulk amount of radioactive wastes

  4. Long term cooling analysis after Fukushima Daiichi accident

    The objective of this study is to analyze of the long term cooling after Fukushima Daiichi accident by RELAP5mode3.3 code and to check the validity of the cooling method. In order to simulate the cooling conditions in Fukushima plants after accident, the model is nodalized on the assumption of the existence of steam/liquid leak position from RPV/PCV and the variety of debris distribution in RPV/PCV. As a result, we estimated the debris distribution in RPV by referring plant parameter such as reactor pressure and temperature. In addition, we performed the analysis of the loss of injection water accident for the current cooling system installed in Fukushima Daiichi cite after the earthquake. In this case, we develop simplified nodalization of RPV to analyze temperature behavior of reactor structural materials by using the radiation heat transfer model. (author)

  5. Reassessment of the technical bases for estimating source terms. Final report

    This document describes a major advance in the technology for calculating source terms from postulated accidents at US light-water reactors. The improved technology consists of (1) an extensive data base from severe accident research programs initiated following the TMI accident, (2) a set of coupled and integrated computer codes (the Source Term Code Package), which models key aspects of fission product behavior under severe accident conditions, and (3) a number of detailed mechanistic codes that bridge the gap between the data base and the Source Term Code Package. The improved understanding of severe accident phenonmena has also allowed an identification of significant sources of uncertainty, which should be considered in estimating source terms. These sources of uncertainty are also described in this document. The current technology provides a significant improvement in evaluating source terms over that available at the time of the Reactor Safety Study (WASH-1400) and, because of this significance, the Nuclear Regulatory Commission staff is recommending its use

  6. Source term code package: modifications and applications

    The Source Term Code Package (STCP) and its development have served a pivotal role in the advancement of source term research by helping to focus both modeling and experimental efforts on issues that are of controlling importance in their ability to predict accident consequences. Currently efforts are continuing with the goal of upgrading various STCP code features. The need for these changes results from the continuing research and growing data base in the area of fission product release and transport. Among the models which are being upgraded are those dealing with direct containment heating, fission product release from the fuel, combined analyses of containment transport, and pool/ice condenser removal, and more direct and adequate coupling of heat transfer and fission product transport in the reactor coolant system

  7. Phase 1 immobilized low-activity waste operational source term

    This report presents an engineering analysis of the Phase 1 privatization feeds to establish an operational source term for storage and disposal of immobilized low-activity waste packages at the Hanford Site. The source term information is needed to establish a preliminary estimate of the numbers of remote-handled and contact-handled waste packages. A discussion of the uncertainties and their impact on the source term and waste package distribution is also presented. It should be noted that this study is concerned with operational impacts only. Source terms used for accident scenarios would differ due to alpha and beta radiation which were not significant in this study

  8. Approximate factorization with source terms

    Shih, T. I.-P.; Chyu, W. J.

    1991-01-01

    A comparative evaluation is made of three methodologies with a view to that which offers the best approximate factorization error. While two of these methods are found to lead to more efficient algorithms in cases where factors which do not contain source terms can be diagonalized, the third method used generates the lowest approximate factorization error. This method may be preferred when the norms of source terms are large, and transient solutions are of interest.

  9. Effects of source term characteristics on Off Site consequence

    Off site consequence analysis in Level 3 PSA is mainly affected by source term release characteristics of nuclear plant. The severe accident analysis codes for quantifying the source term release characteristics such as MELCOR and MAAP provide detailed information of these characteristics to assess the off site consequence. The aforementioned characteristics, however, have not been considered in the consequence analysis of domestic plants because of large uncertainty in these characteristics so far. Recently, the USNRC SOARCA report showed an approach to utilize detailed source term characteristics provided by MELCOR code to quantify the off site consequence more realistically. Main purpose of this study is to assess effects of the MELCOR source term characteristics on the off site consequence analysis of a domestic nuclear power plant, in a similar fashion to the SOARCA approach. Among many features characterizing source term, the most important one is to determine initial and boundary conditions of atmospheric dispersion such as:- Release amounts of source term - Release time and duration Moreover, plumes features (i.e., radiation clouds) affect atmospheric dispersion that shapes plume characteristics as follows: - Initial dimension of plumes - Plume rise - Deposition of radioactive materials during dispersion Although the current severe accident codes have some limitation in providing the entire source term release characteristics needed in the consequence analysis, the essential information for these features could be obtained from these codes. It is noted that the typical approaches, which generate source term information for the consequence analysis from the severe accident codes, should require a technical manipulation by the experts of consequence analysis. The present effort focused on an identification of insights to utilize source term characteristics of the severe accident codes

  10. Procedures: Source Term Measurement Program

    The report contains procedures for the Source Term Measurement Project being performed by Idaho National Engineering Laboratory for the Nuclear Regulatory Commission. This work is being conducted for the Office of Nuclear Regulatory Research in support of requirements of the Effluent Treatment Systems Branch of the Office of Nuclear Reactor Regulation. This project is designed to obtain source term information at operating light water reactors to update the parameters used in NRC calculational models (GALE codes). Detailed procedures and methods used for collection and analysis of samples are presented. This provides a reference base to supplement a series of reports to be issued by the Source Term Measurements Project which will present data obtained from measurements in specific nuclear power stations. Reference to appropriate parts of these procedures will be made as required

  11. Source term analyses for the Muehleberg nuclear power plant

    In the study presented here, the source terms for six accident scenarios at the Muehleberg nuclear power station were investigated; namely two low pressure incidents, a high pressure incident, a fire, an earthquake and a plane crash. figs., tabs., 44 refs

  12. Effect of source term composition on offsite doses

    The development of new realistic accident source terms has identified the need to establish a basis for comparing the impact of such source terms. This paper attempts to develop a generalized basis of comparison by investigating contributions to offsite acute whole body doses from each group of radionuclides being released to the atmosphere, using CRAC2. The paper also investigates the effect of important parameters such as regional meteorology, sheltering, and duration of release. Finally, the paper focuses on significant changes in the relative importance of individual radionuclide groups in PWR2, SST1, and a revision of the Stone and Webster proposed interim source term

  13. Thermochemical considerarations in source term evaluation

    The calculation of the release of fission products from degraded fuel in a light water reactor core uncovery accident is the first step in determining the overall radiological source term. It is the aim of costly experiments to improve our knowledge about the release behavior of the relevant fission products. Since this depends greatly on their chemistry, a thermodynamic evaluation about compound formation and vaporisation in a fuel-fission product-coolant system should precede such tests. It shows how the volatility of these products may change with test conditions. It will need more reduction of the steam atmosphere to get a noticable release of barium and strontium than to have europium show up. It is very unlikely that ruthenium is significantly released even in a nonreduced steam environment. Molybdenum will be released with the cesium in oxidising and slightly reducing atmospheres. Boron has an effect on the iodine and cesium chemistry. This, however, depends greatly on test or accident conditions. It is practically nonexistent at high steam pressures. Low oxygen potential and high boron content of the atmosphere increase the effect. (orig.)

  14. Technology Development on Alternate Source Term Analysis and Application

    Source term presented in TID-14844 and Regulatory Guide 1.4 has been used to estimate radiation dose from design basis accidents. However, a more realistic and physically-based source term, alternate source term, has been developed and presented in NUREG-1465 and Regulatory Guide 1.183. In addition, the concept and criteria of radiological dose estimation has been changed through the serial publications of ICRP-9, 26, and 60. In ICRP-60, ICRP introduced effective dose concept in stead of dose concept based on critical organ and whole body since the publication of ICRP-9. Korean regulatory authority is planning to issue the new regulation to adopt the alternate source term and the effective dose concept to radiation dose analysis for design basis accidents. As a measure for the issuance of the new regulation, the application methodology of alternate source term and effective dose for design basis accidents was established and merged to the computer program called DBADOSE. This program was verified in accordance with the verification procedure. The result of calculation by DBADOSE showed small difference of less than 5% in comparison with the result of STARDOSE which was developed by Polestar Applied Technology, Inc. Kori Units 3,4 was selected as a pilot plant to apply the alternate source term and the effective dose. The margins to licensing criteria were reanalyzed for design basis accidents. As a result of this application, it was assured that current design of Kori Units 3,4 has enough margins and design simplification were proposed. (authors)

  15. Evaluation of released source terms from burning mock combustible waste

    To evaluate quantitatively confinement capability of the radioactive materials in the nuclear fuel facility under the fire accident, analysis of accident sequence, including clogging characteristics of the ventilation filters, needs to be performed. For the purpose of the evaluation, accumulation of the source term data such as release rates of the smoke and energy, and particle size distribution of the smoke during the fire accident is necessary. Therefore, experiments for evaluating burning characteristics of combustible solid wastes and recovered solvents, which are disposed from the facilities, have been performed by using the mock combustible wastes and the method for estimating the source terms has been investigated. When mixtures of rubber and cloth gloves as mock combustible solid wastes were burnt, the smoke with above 1 μm in diameter was confined in the carbonized residue of cloth gloves and the release ratio of the smoke in the burning of mixtures was decreased compared with the burning of only rubber gloves. The source terms were evaluated with the cell ventilation system safety analysis code CELVA-1D by using the experimental results as the input, such as temperature of the gas phase, total burnt weight and total collected weight of the smoke under the burning of rubber gloves as mock wastes. The source terms calculated by the CELVA-1D reasonably agreed with the values estimated from the recommended calculation parameters in the Nuclear Fuel Cycle Facility Accident Analysis Handbook (NUREG-1320). Therefore, the present CELVA-1D method for evaluating the source terms during burning is considered to be valid. This means that the source terms can be estimated by using this method if the information such as the temperature of the gas phase, total burnt weight and total collected weight of the smoke are given. (author)

  16. Phenomena of the ex-vessel source term

    The release of radionuclides and the production of aerosols from core debris outside the vessel in a severe reactor accident are discussed. This ex-vessel source term is caused by expulsion of core debris from a pressurized vessel, core debris interactions with water and core debris interactions with concrete. The models and data available for the ex-vessel source term are described and compared with the models used in the Reactor Safety Study (WASH-1400). The release associated with core debris interactions with concrete has been investigated extensively. Current perceptions concerning the release during these interactions are quantitatively and qualitatively different than perceptions arising from the Reactor Safety Study model. Source terms arising from other ex-vessel phenomena have not been studied extensively, yet these source terms may be important. There is a great need for experimental data for all aspects of the ex-vessel source term. (author)

  17. Unconventional sources of plant information for accident management

    An essential element to accident management is having as clear a picture as is practical of the plant status and thus of the accident and its progress. Effective, appropriate decisions to control and mitigate an accident are dependent on making this assessment of the accident. The objective of this paper is to stimulate consideration of unconventional plant information sources through discussion of specific examples. A plant's condition during an accident can be characterized by plant parameters such as temperatures and pressures and by plant system operational status. For example, core damage is associated with increasing temperatures, pressures, and radiation levels in many different systems and plant areas. Reg. Guide 1.97 instrumentation exists to provide information to allow operators to take specified manual actions (Type A), to indicate whether plant safety functions are being accomplished (Type B), to indicate the potential for breach of barriers to fission product release (Type C), to indicate operability of individual safety systems (Type D), and to indicate the magnitude of radioactive material releases (Type E). Reg. Guide 1.97 instrument range requirements, with the exception of pressure instruments, address conditions up to design basis conditions. Pressure instrument range requirements exceed design basis conditions. During a severe accident, some instruments may not see conditions beyond their design basis. Effective accident management includes the ability to establish a consistent picture of the accident by accumulating information from as many sources as is practical. Operability of systems and components, and non-safety related temperature, radiation, pressure, and water-level indication can be used to directly indicate, measure, or infer plant parameters which confirm, augment or replace those otherwise available. Innovative uses of information sources thus serve to increase the diversity and flexibility of accident data available. Both the

  18. Long term health effects in Sweden from the Chernobyl accident

    The morning of 28 April 1986 was the beginning of an intensive period of radiation protection work in Sweden. During that morning the Chernobyl accident became known in the western world through the detection of radioactive contamination in Sweden and at the Forsmark nuclear power plant in particular. The environmental consequences of the fallout have been studied in various research projects. The effects on agriculture in Sweden was mainly limited to the first year after the accident. The long term effects are instead seen in products from the semi-natural ecosystems: in moose, roedeer, reindeer, mushrooms and fish from lakes in areas with a high deposition of radioactive caesium. High concentrations of 137Cs in reindeer meat in combination with an estimated effective ecological half-life of about 4 years, will cause problems for reindeer husbandry in the most contaminated parts for many years to come. In moose, roedeer and mushrooms, the ecological half-lives are very long and in some compartments seem to approach the physical half-life of 137Cs. 22 refs, 3 figs

  19. Source terms due to the activated corrosion products in primary cooling loops of ITER

    The paper deals with the Source Terms due to activated corrosion products escaping from a primary cooling loop of the ITER blanket following a Loss Of Cooling Accident (LOCA). Both in-vessel and out of vessel accident are considered. The assessment is based on the European multi-code methodological approach set-up to estimate the environmental releases of the activated corrosion/erosion products involved in the accident scenarios of a fusion machine. The approach is based on the following parameters: Radioactivity Inventory (RI), Process Source Terms (PST), and Environmental Source Terms (EST). Different codes have been used and compared to evaluate such parameters

  20. Atucha-I source terms for sequences initiated by transients

    The present work is part of an expected source terms study in the Atucha I nuclear power plant during severe accidents. From the accident sequences with a significant probability to produce core damage, those initiated by operational transients have been identified as the most relevant. These sequences have some common characteristics, in the sense that all of them resume in the opening of the primary system safety valves, and leave this path open for the coolant loss. In the case these sequences continue as severe accidents, the same path will be used for the release of the radionuclides, from the core, through the primary system and to the containment. Later in the severe accident sequence, the failure of the pressure vessel will occur, and the corium will fall inside the reactor cavity, interacting with the concrete. During these processes, more radioactive products will be released inside the containment. In the present work the severe accident simulation initiated by a blackout is performed, from the point of view of the phenomenology of the behavior of the radioactive products, as they are transported in the piping, during the core-concrete interactions, and inside the containment buildings until it failure. The final result is the source term into the atmosphere. (author)

  1. Long-term followup of patients involved in radiation accidents

    This paper discusses how followup of patients involved in accidental exposures should be tailored to the circumstances of the accident. The critical issues for long-term followup are divided into analysis at relatively low absorbed doses for carcinogenic effects, and such followup may take the form of epidemiologic studies that may need to be continued over a period of decades. With higher doses, direct or nonstochastic effects are important, and the exact followup scheme that should be utilized depends upon the patient, absorbed doses and tissues irradiated. In general, patients exceeding the REAC/TS guidelines for significant exposure are followed using the Navy protocol (NAVmed). Certainly, for significant exposures, appropriate medical consultation in design of the followup procedure is preferable to a routine protocol

  2. Fission product source terms and engineered safety features

    New, technically defensible, methodologies to establish realistic source term values for nuclear reactor accidents are discussed. Although these methodologies will undoubtedly find widespread use in the development of emergency response procedures, that is, procedures to be implemented external to the plant, such as sheltering or evacuation of the surrounding population, it is less clear that the industry is preparing to employ the newer results to develop a more rational approach to the implementation of engineered safety features for the mitigation of fission product releases in the event of a nuclear reactor accident

  3. Survey of source term codes

    The type, scope and role of current source term codes is reviewed. Issues and problems that are generic to most or all such codes are identified. These include numerical problems, fundamental physical and chemical databases, uncertainties in fuel degradation and thermal-hydraulics calculations, how to benchmark systems application codes against mechanistic codes and weaknesses in phenomenological modelling. It is concluded that further code development to address these issues is justified. This should be tightly coupled to experimental work within an internationally agreed framework. (Author)

  4. Source term calculations for assessing radiation dose to equipment

    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 accident have been developed. The Surry reactor has been selected in this study as representative of PWR containment designs. Similarly, the Peach Bottom reactor has been used to examine radionuclide distributions in boiling water reactors. The time-dependent inventory of each key radionuclide is provided in terms of its activity in curies. The data are to be used by Sandia National Laboratories to perform shielding analyses to estimate radiation dose to equipment in each containment design. See NUREG/CR-5175, ''Beta and Gamma Dose Calculations for PWR and BWR Containments.'' 6 refs., 11 tabs

  5. Literature study of source term research for PWRs

    A literature survey has been carried out in support of ongoing source term calculations with the MELCOR code of some severe accident scenarios for the Swedish Ringhals 2 pressurised water reactor (PWR). The research in the field of severe accidents in power reactors and the source term for subsequent release of radioisotopes was intensified after the Harrisburg accident and has produced a large amount of reports and papers. This survey was therefore limited to research concerning PWR type of reactors and with emphasis on papers related to MELCOR code development. A background is given, relating to some historic documents, and then more recent research after 1990 is reviewed. Of special interest is the ongoing PMbus-programme which is creating new and important results of benefit to the code development and validation of, among others, the MELCOR code. It is concluded that source term calculations involve simulation of many interacting complex physical phenomena, which result in large uncertainties The research has, however, over the years led to considerable improvements Thus has the uncertainty in source term predictions been reduced one to two orders of magnitude from the simpler codes in the early 1980-s to the more realistic codes of today, like MELCOR

  6. Literature study of source term research for PWRs

    Sponton, L.L.; NiIsson, Lars

    2001-04-01

    A literature survey has been carried out in support of ongoing source term calculations with the MELCOR code of some severe accident scenarios for the Swedish Ringhals 2 pressurised water reactor (PWR). The research in the field of severe accidents in power reactors and the source term for subsequent release of radioisotopes was intensified after the Harrisburg accident and has produced a large amount of reports and papers. This survey was therefore limited to research concerning PWR type of reactors and with emphasis on papers related to MELCOR code development. A background is given, relating to some historic documents, and then more recent research after 1990 is reviewed. Of special interest is the ongoing PMbus-programme which is creating new and important results of benefit to the code development and validation of, among others, the MELCOR code. It is concluded that source term calculations involve simulation of many interacting complex physical phenomena, which result in large uncertainties The research has, however, over the years led to considerable improvements Thus has the uncertainty in source term predictions been reduced one to two orders of magnitude from the simpler codes in the early 1980-s to the more realistic codes of today, like MELCOR.

  7. Source term experiments program (STEP)

    Four experiments were conducted in the TREAT facility to investigate the behaviour of fission products released from typical LWR fuel overheated to the point of catastrophic cladding degradation. Heatup and steam flow transients were used that simulate the conditions expected in operating power reactors undergoing various types of hypothetical severe accidents. The experiments were integral in nature and aimed at the physico-chemical characterization, near the point of origin, of the biologically important volatile fission products released early in such accidents. Programme strategy and experimental methods are discussed briefly, and the features of test operations and preliminary results are presented in more detail. (author)

  8. Source Term Experiments Program (STEP)

    Four experiments were conducted in the TREAT facility to investigate the behavior of fission products released from typical LWR fuel overheated to the point of catastrophic cladding degradation. Heatup and steam flow transients were used that simulate the conditions expected in operating power reactors undergoing various types of hypothetical severe accidents. The experiments were integral in nature and aimed at the physicochemical characterization, near the point of origin, of the biologically important volatile fission products released early in such accidents. Program strategy and experimental methods are discussed briefly, and features of test operations and preliminary results are presented in more detail

  9. In-Plant Fission Product Behavior in SGTR Accident with Long-Term SBO

    The off-site AC power was recovered in 9 days after the accident in the NPS. Therefore safety injection by fire pump truck with fresh water or seawater is only available in the Fukushima accident. However, safety injection by fire pump truck is not always effective due to the high pressure of RPV inside or leakages of alternative water injection flow paths. In the SBO situations in pressurized water reactor plant (PWR), turbine driven auxiliary feedwater (TD-AFW) pump can inject water to the secondary side of steam generator. However, turbine inlet steam flow control valve cannot work properly when loss of vital DC power occurs. Vital DC power is designed to be maintained during 4 or 8 hours in the SBO conditions. In this paper motor-driven and turbine driven AFW pumps are all assumed to be not working at time 0 sec as a worst case assumption. Iodine pool-scrubbing can occur in the secondary side of the faulted steam generator. However, iodine pool-scrubbing in the secondary side of the faulted steam generator is assumed not to be working, due to the assumption of the loss of DC battery for turbine inlet flow control valve. Iodine pool-scrubbing is one of the long-term research issues in safety assessment of nuclear power plant severe accident. PHEBUS FPT series and THAI experiment projects are typical projects on the resolving source term issues in severe accident of nuclear power plants. However, iodine retention by pool scrubbing is still a debating issue. In such containment bypass sequences, fission products can be released out to environment directly from RCS without retention or deposition in containment structures. SGTR is one of the hazardous accident scenarios in the typical PSA, because SGTR induces a large release amount of source term to environment directly. A key operation strategy was the isolation of the broken reactor coolant system loop from the intact loop. Typical core degradation in SGTR scenarios occurs with multiple failures of the isolation

  10. In-Plant Fission Product Behavior in SGTR Accident with Long-Term SBO

    Kim, Tae Woon; Han, Seok Jung; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The off-site AC power was recovered in 9 days after the accident in the NPS. Therefore safety injection by fire pump truck with fresh water or seawater is only available in the Fukushima accident. However, safety injection by fire pump truck is not always effective due to the high pressure of RPV inside or leakages of alternative water injection flow paths. In the SBO situations in pressurized water reactor plant (PWR), turbine driven auxiliary feedwater (TD-AFW) pump can inject water to the secondary side of steam generator. However, turbine inlet steam flow control valve cannot work properly when loss of vital DC power occurs. Vital DC power is designed to be maintained during 4 or 8 hours in the SBO conditions. In this paper motor-driven and turbine driven AFW pumps are all assumed to be not working at time 0 sec as a worst case assumption. Iodine pool-scrubbing can occur in the secondary side of the faulted steam generator. However, iodine pool-scrubbing in the secondary side of the faulted steam generator is assumed not to be working, due to the assumption of the loss of DC battery for turbine inlet flow control valve. Iodine pool-scrubbing is one of the long-term research issues in safety assessment of nuclear power plant severe accident. PHEBUS FPT series and THAI experiment projects are typical projects on the resolving source term issues in severe accident of nuclear power plants. However, iodine retention by pool scrubbing is still a debating issue. In such containment bypass sequences, fission products can be released out to environment directly from RCS without retention or deposition in containment structures. SGTR is one of the hazardous accident scenarios in the typical PSA, because SGTR induces a large release amount of source term to environment directly. A key operation strategy was the isolation of the broken reactor coolant system loop from the intact loop. Typical core degradation in SGTR scenarios occurs with multiple failures of the isolation

  11. Inventory and source term evaluation of Russian nuclear power plants for marine applications

    This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

  12. Inventory and source term evaluation of Russian nuclear power plants for marine applications

    Reistad, O. [Norwegian Radiation Protection Authority (Norway); Oelgaard, P.L. [Risoe National Lab. (Denmark)

    2006-04-15

    This report discusses inventory and source term properties in regard to operation and possible releases due to accidents from Russian marine reactor systems. The first part of the report discusses relevant accidents on the basis of both Russian and western sources. The overview shows that certain vessels were much more accident prone compared to others, in addition, there have been a noteworthy reduction in accidents the last two decades. However, during the last years new types of incidents, such as collisions, has occurred more frequently. The second part of the study considers in detail the most important factors for the source term; reactor operational characteristics and the radionuclide inventory. While Russian icebreakers has been operated on a similar basis as commercial power plants, the submarines has different power cyclograms which results in considerable lower values for fission product inventory. Theoretical values for radionuclide inventory are compared with computed results using the modelling tool HELIOS. Regarding inventory of transuranic elements, the results of the calculations are discussed in detail for selected vessels. Criticality accidents, loss-of-cooling accidents and sinking accidents are considered, bases on actual experiences with these types of accident and on theoretical considerations, and source terms for these accidents are discussed in the last chapter. (au)

  13. Selected source term topics. Report to CSNI by an OECD/NEA Group of experts

    CSNI Report 136 summarizes the results of the work performed by the Group of Experts on the Source Term and Environmental Consequences (PWG4) during the period extending from 1983 and 1986. This report is complementary to Part 1, 'Technical Status of the Source Term' of CSNI Report 135, 'Report to CSNI on Source Term Assessment, Containment atmosphere control systems, and accident consequences'; it considers in detail a number of very specific issues thought to be important in the source term area. It consists of: an executive summary (prepared by the Chairman of the Group), a section on conclusions and recommendations, and five technical chapters (fission product chemistry in the primary circuit of a LWR during severe accidents; resuspension/re-entrainment of aerosols in LWRs following a meltdown accident; iodine chemistry under severe accident conditions; effects of combustion, steam explosions and pressurized melt ejection on fission product behaviour; radionuclide removal by pool scrubbing), a technical annex and two appendices

  14. Considerations about source term now used aiming to emergency planning

    The applicability of source terms, in parametric studies for improving external emergengy plan for Angra-I reactor is presented. The source term is defined as, the quantity and radioactive material disposable for releasing to the environment in case of austere accident in a nuclear power plant. The following hypothesis: occuring accident, 100% of the noble gases, 50% of halogens and 1% of solid fission products contained into the reactor core, are released immediately toward the containment building; the radioactivity releasing to the environment is done at a constant rate of 0.1% in mass per day; the actuation of mitigated systems of radioactivity releasing, such as, spray of container or system of air recirculation by filters, is not considered; and the releasing is done at soil level. (M.C.K.)

  15. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an atmospheric dispersion model with an improved deposition scheme and oceanic dispersion model

    G. Katata; Chino, M; T. Kobayashi; Terada, H.; Ota, M; Nagai, H.; M. Kajino; Draxler, R; M. C. Hort; Malo, A.; Torii, T.; Y. Sanada

    2015-01-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measure...

  16. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

    G. Katata; Chino, M; T. Kobayashi; Terada, H.; Ota, M; Nagai, H.; M. Kajino; Draxler, R; M. C. Hort; Malo, A.; Torii, T.; Y. Sanada

    2014-01-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSP...

  17. Unconventional sources of plant information for accident management

    Oehlberg, R. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); Machiels, A. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); Chao, J. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); Weiss, J. (Electric Power Research Inst., Palo Alto, CA (United States). Nuclear Power Div.); True, D. (ERIN Engineering and Research, Inc., Walnut Creek, CA (United States)); James, R. (ERIN Engineering and Research, Inc., Walnut Creek, CA (United States))

    1992-07-01

    The paper highlighted that other information sources can help to augment and confirm data available from dedicated accident instrumentation such as Reg. Guide 1.97 Instrumentation: inferences of plant status are possible from measurements and measurement trends obtained from instruments not expected to function, observations of system or component operability/inoperability, and observations of locally harsh environmental conditions. Detailed plant-specific examples are given, e.g. regarding the reactor pressure and level indication in BWRs, or the reactor cavity temperature indication on WE-type PWRs which the authors speculate may yield information related to vessel and core temperature. The authors advocate that others look at their information sources in a creative way. (orig.)

  18. Impact of short-term severe accident management actions in a long-term perspective. Final Report

    The present systems for severe accident management are focused on mitigating the consequences of special severe accident phenomena and to reach a safe plant state. However, in the development of strategies and procedures for severe accident management, it is also important to consider the long-term perspective of accident management and especially to secure the safe state of the plant. The main reason for this is that certain short-term actions have an impact on the long-term scenario. Both positive and negative effects from short-term actions on the accident management in the long-term perspective have been included in this paper. Short-term actions are accident management measures taken within about 24 hours after the initiating event. The purpose of short-term actions is to reach a stable status of the plant. The main goal in the long-term perspective is to maintain the reactor in a stable state and prevent uncontrolled releases of activity. The purpose of this short Technical Note, deliberately limited in scope, is to draw attention to potential long-term problems, important to utilities and regulatory authorities, arising from the way a severe accident would be managed during the first hours. Its objective is to encourage discussions on the safest - and maybe also most economical - way to manage a severe accident in the long term by not making the situation worse through inappropriate short-term actions, and on the identification of short-term actions likely to make long-term management easier and safer. The Note is intended as a contribution to the knowledge base put at the disposal of Member countries through international collaboration. The scope of the work has been limited to a literature search. Useful further activities have been identified. However, there is no proposal, at this stage, for more detailed work to be undertaken under the auspices of the CSNI. Plant-specific applications would need to be developed by utilities

  19. Severe accident risk minimization studies for the Advanced Neutron Source (ANS) reactor plant at the Oak Ridge National Laboratory

    This paper discusses salient aspects of severe accident related phenomenological considerations, scoping studies, and mitigative design features being studied for incorporation into a high-power research reactor plant. Key results of scoping studies on steam explosions, recriticality, core-concrete interactions, and containment transport are highlighted. Evolving design features of the containment are described. Containment response calculations for a site-suitability basis transient are presented that demonstrate acceptable source term values and superior containment performance. Oak Ridge National Laboratory's (ORNL) Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management initiated severe accident analysis and related technology development early-on in the design phase itself. This was done to aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It also provides a means for satisfying on- and off-site regulatory requirements, accident-related dose exposures, and containment response and source-term best-estimate analyses for level-2 and -3 Probabilistic Risk Analysis (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions and consequently provide insights for the development of strategies and design philosophies for accident mitigation, management, and emergency preparedness efforts

  20. Summary of major accidents with radiation sources and lessons learned

    The paper reviews some of the major radiological accidents that have occurred around the world and identifies key lessons to be learned from them. It emphasizes the value of feedback from the reporting of accidents, the need for effective reporting mechanisms and, most important, the importance of acting on the lessons learned to ensure accident prevention. (author)

  1. Analysis of a BWR MKI utilizing advanced BMI-2104 computational techniques to calculate source terms

    The effects of incorporating code modifications to more realistically represent accident phenomenology were examined with specific focus on the BWR MKI AE γ-sequence. Results indicate that the source terms are only a fraction of those reported in BMI-2104 and that core/concrete interaction is not a significant source term contributor for the BWR MKI containments. (author)

  2. Application of the source term code package to obtain a specific source term for the Laguna Verde Nuclear Power Plant

    The main objective of the project was to use the Source Term Code Package (STCP) to obtain a specific source term for those accident sequences deemed dominant as a result of probabilistic safety analyses (PSA) for the Laguna Verde Nuclear Power Plant (CNLV). The following programme has been carried out to meet this objective: (a) implementation of the STCP, (b) acquisition of specific data for CNLV to execute the STCP, and (c) calculations of specific source terms for accident sequences at CNLV. The STCP has been implemented and validated on CDC 170/815 and CDC 180/860 main frames as well as on a Micro VAX 3800 system. In order to get a plant-specific source term, data on the CNLV including initial core inventory, burn-up, primary containment structures, and materials used for the calculations have been obtained. Because STCP does not explicitly model containment failure, dry well failure in the form of a catastrophic rupture has been assumed. One of the most significant sequences from the point of view of possible off-site risk is the loss of off-site power with failure of the diesel generators and simultaneous loss of high pressure core spray and reactor core isolation cooling systems. The probability for that event is approximately 4.5 x 10-6. This sequence has been analysed in detail and the release fractions of radioisotope groups are given in the full report. 18 refs, 4 figs, 3 tabs

  3. Source-term reevaluation for US commercial nuclear power reactors: a status report

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date

  4. Source-term reevaluation for US commercial nuclear power reactors: a status report

    Herzenberg, C.L.; Ball, J.R.; Ramaswami, D.

    1984-12-01

    Only results that had been discussed publicly, had been published in the open literature, or were available in preliminary reports as of September 30, 1984, are included here. More than 20 organizations are participating in source-term programs, which have been undertaken to examine severe accident phenomena in light-water power reactors (including the chemical and physical behavior of fission products under accident conditions), update and reevaluate source terms, and resolve differences between predictions and observations of radiation releases and related phenomena. Results from these source-term activities have been documented in over 100 publications to date.

  5. Agricultural production as a source of irradiation of populations in radiation accidents

    Radioactive contamination of the agricultural production sphere in case of a radiation accident with radionuclide release is one of the most important sources of additional irradiation for population. As a result, realisation of the measures for mitigating the consequences of the accident in agro-industrial complex assumes the leading role in total system of measures providing radiation safety. The possibility to obtain agricultural production meeting the radiological standards is one of the main indications of effectiveness of liquidation of the accident consequences. First, obtaining of agricultural production meeting the radiological standards provides decreasing of the total radiation dose. The evidence is obtained in the 8-year experience of liquidation of the Chernobyl NPP accident that the 70% decrease of the internal dose for population became possible through a complex of protection measures in agriculture (during the first year after the accident, the contribution of internal dose into the total one amounted to 45%, and that of external irradiation - 52%; for the 70-year period these values are 39% and 60%, respectively). Second, the possibility to obtain 'pure' agricultural production is one of the most important factors of psychological stability for population. Third, obtaining of consumable (as to radionuclide content) agricultural production in private small holdings is one of the guarantees of stability of demographic sector in the accident-affected zone. From the point of view of organization of agricultural production in liquidation of the consequences of accidents with radioactive releases into environment, some periods can be distinguished: (i) the first (early) period takes 10-12 days after the accident. The main measures in the field of agricultural production are in operative assessment of the radiological situation, organization of radiation survey, express classification of agricultural products ready to consumption. If radionuclide content

  6. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    1978-02-01

    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 deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

  7. HTGR Mechanistic Source Terms White Paper

    Wayne Moe

    2010-07-01

    The primary purposes of this white paper are: (1) to describe the proposed approach for developing event specific mechanistic source terms for HTGR design and licensing, (2) to describe the technology development programs required to validate the design methods used to predict these mechanistic source terms and (3) to obtain agreement from the NRC that, subject to appropriate validation through the technology development program, the approach for developing event specific mechanistic source terms is acceptable

  8. Estimated long term health effects of the Chernobyl accident

    Cardis, E. [International Agency for Research on Cancer, Lyon (France)

    1996-07-01

    Apart from the dramatic increase in thyroid cancer in those exposed as children, there is no evidence to date of a major public health impact as a result of radiation exposure due to the Chernobyl accident in the three most affected countries (Belarus, Russia, and Ukraine). Although some increases in the frequency of cancer in exposed populations have been reported ,these results are difficult to interpret, mainly because of differences in the intensity and method of follow-up between exposed populations and the general population with which they are compared. If the experience of the survivors of the atomic bombing of Japan and of other exposed populations is applicable, the major radiological impact of the accident will be cases of cancer. The total lifetime numbers of excess cancers will be greatest among the `liquidators` (emergency and recovery workers) and among the residents of `contaminated` territories, of the order of 2000 to 2500 among each group (the size of the exposed populations is 200,000 liquidators and 3,700,000 residents of `contaminated` areas). These increases would be difficult to detect epidemiologically against an expected background number of 41500 and 433000 cases of cancer respectively among the two groups. The exposures for populations due to the Chernobyl accident are different in type and pattern from those of the survivors of the atomic bombing of Japan. Thus predictions derived from studies of these populations are uncertain. The extent of the incidence of thyroid cancer was not envisaged. Since only ten years have lapsed since the accident, continued monitoring of the health of the population is essential to assess the public health impact.

  9. RAMA Sources term group final report January 1985

    This report briefly summarizes the phenomena believed to occur during a severe core meltdown accident and discusses their chemical and physical significance for the release, transport and deposition of radioactive materials during such accidents. The models used in the computer codes FPRAT, AEROREL, MATORET, RETAIN and NUCLEIDS are described and discussed in light of current knowledge and the experience obtained by running these codes. Possible deficiencies in the models are identified and some suggestions are given for further work on code development. Some results from a limited sensitivity analysis and from case tests with the codes are also presented. Finally the report presents the conclusions drawn by the RAMA source term group regarding the usefulness of, correctness of results from and desired further development of the codes tested and used within the RAMA project. (author)

  10. Influence of iodine chemistry on source term assessment

    The major goal of a phenomenology analysis of containment during a severe accident situation can be spitted into the following ones: to know the containment response to the different loads and to predict accurately the fission product and aerosol behavior. In this report, the main results coming from the study of a hypothetical accident scenario, based on LA-4 experiment of LACE project, are presented. In order to do it, several codes have been coupled: CONTEMPT4/MOD5 (thermohydraulics), NAUA/MOD5 (aerosol physics) and IODE (iodine chemistry). It has been demonstrated the impossibility of assessing with confidence the Source Term if the chemical conduct of some radionuclides is not taken into account. In particular, the influence on the iodine retention efficiency of the sump of variables such as pH has been proven. (Author)12 refs

  11. Running the source term code package in Elebra MX-850

    The source term package (STCP) is one of the main tools applied in calculations of behavior of fission products from nuclear power plants. It is a set of computer codes to assist the calculations of the radioactive materials leaving from the metallic containment of power reactors to the environment during a severe reactor accident. The original version of STCP runs in SDC computer systems, but as it has been written in FORTRAN 77, is possible run it in others systems such as IBM, Burroughs, Elebra, etc. The Elebra MX-8500 version of STCP contains 5 codes:March 3, Trapmelt, Tcca, Vanessa and Nava. The example presented in this report has taken into consideration a small LOCA accident into a PWR type reactor. (M.I.)

  12. Calculation and analysis of radioactive source term in PWR assemblies

    Background: When fission occurs in fuel of reactor core, it produces a large amount of radioactive materials, which may cause harm to the environment and human health. Purpose: The radioactive materials in fuel could provide input data for shielding design of reactor coolant radioactive source term, analysis of accident source term and radioactive consequence assessment. Methods: The calculation of radioactive source in fuel was studied for pressurized water reactor: the calculation methods and models were established using ORIGEN-S, and the difference of nuclides radioactivity under different burnup was also studied. The effect of different versions of ENDF/B cross-section database on the calculation results was analyzed, so as to provide a basis for the calculation of radioactive source in fuel. Results: The results showed that the method established by ORIGEN-ARP was more suitable for calculating radioactive source term in fuel assemblies and the different versions of ENDF/B database had a great impact on radioactivity calculation. Conclusion: Based on the ENDF/B-VII database, using ORIGEN-ARP to calculate radioactive source term in fuel assemblies could not only improve efficiency, but also improve the calculation accuracy. (authors)

  13. Strengthening long term control over radioactive sources

    The traditional focus of the regulation of radioactive sources is the protection of workers and the public from the misuse of sources and from accidents. Security measures were also a concern, but with the principal aim of preventing petty theft or accidental loss. Our concern, of course, is that a high risk radioactive source might be married with conventional explosives and used in a radiological dispersal device (RDD). Means must be found to protect the public from the use of high risk radioactive sources in an RDD. The task may appear daunting at first because of the widespread use of radioactive sources throughout the world. Compounding the problem is the fact that there also is a general lack of effective domestic controls on even high risk radioactive sources. The IAEA has noted that more than 100 countries lack effective control over radiation sources because most do not have the required infrastructure. The US Nuclear Regulatory Commission (NRC), like its counterparts in other countries, has found that modification of our regulatory programme to account for the terrorist threat is necessary. Although the work on this problem is still under way, some of the components are underlined that are believed to be the elements of an effective regulatory programme to counteract the RDD threat. The aim is a programme that achieves an appropriate balance of safety, security, public benefit and economic feasibility. The main objectives covered in this presentation cover the following topics: Categorization; Security measures; Imports/exports; Disposal; Orphan sources; Emergency response

  14. Source term survey for the source compartment of a ray instrument plant

    In April 2011, an investigation of source term was performed for the radioactive source compartment of a ray instrument plant, being set to be decommissioned. The results showed that gamma radiation air absorbed dose rate and surface contamination monitoring values were relatively high in the wall and ceiling near the cobalt source of calibration chamber 1 and the values were at the normal background level in the other area of the compartment. The water in the well was not contaminated by radiation materials. 5 radioactive sources in the source compartment were detected with the total activity of 2.54×1011 Bq. Among which 3 radioactive sources were reported for the first time with a activity level of 2.99×109 Bq and then the potential radiation accident can be avoided because of the survey. (authors)

  15. Modified ensemble Kalman filter for nuclear accident atmospheric dispersion: Prediction improved and source estimated

    Highlights: • A modified ensemble Kalmen filter data assimilation method is proposed. • The method can consider four main uncertain parameters in the puff model. • The prediction of radioactive material atmospheric dispersion is improved. • The source release rate and plume rise height are successfully reconstructed. • It can shorten the time lag in the response of ensemble Kalmen filter. - Abstract: Atmospheric dispersion models play an important role in nuclear power plant accident management. A reliable estimation of radioactive material distribution in short range (about 50 km) is in urgent need for population sheltering and evacuation planning. However, the meteorological data and the source term which greatly influence the accuracy of the atmospheric dispersion models are usually poorly known at the early phase of the emergency. In this study, a modified ensemble Kalman filter data assimilation method in conjunction with a Lagrangian puff-model is proposed to simultaneously improve the model prediction and reconstruct the source terms for short range atmospheric dispersion using the off-site environmental monitoring data. Four main uncertainty parameters are considered: source release rate, plume rise height, wind speed and wind direction. Twin experiments show that the method effectively improves the predicted concentration distribution, and the temporal profiles of source release rate and plume rise height are also successfully reconstructed. Moreover, the time lag in the response of ensemble Kalman filter is shortened. The method proposed here can be a useful tool not only in the nuclear power plant accident emergency management but also in other similar situation where hazardous material is released into the atmosphere

  16. PST - a new method for estimating PSA source terms

    The Parametric Source Term (PST) code has been developed for estimating radioactivity release fractions. The PST code is a framework of equations based on activity transport between volumes in the release pathway from the core, through the vessel, through the containment, and to the environment. The code is fast-running because it obtains exact solutions to differential equations for activity transport in each volume for each time interval. It has successfully been applied to estimate source terms for the six Pressurized Water Reactors (PWRs) that were selected for initial consideration in the Accident Sequence Precursor (ASP) Level 2 model development effort. This paper describes the PST code and the manner in which it has been applied to estimate radioactivity release fractions for the six PWRs initially considered in the ASP Program

  17. Quantities of I-131 and Cs-137 in accumulated water in the basements of reactor buildings in process of core cooling at Fukushima Daiichi nuclear power plants accident and its influence on late phase source terms

    During the process of core cooling at Fukushima Daiichi nuclear power plants accident, large amount of contaminated water was accumulated in the basements of the reactor buildings at Units 1-4. The present study estimated the quantities of I-131 and Cs-137 in the water as of late March based on the press-opened data. The estimated ratios of I-131 and Cs-137 quantities to the core inventories are 0.51%, 0.85% at Unit 1, 74%, 38% at Unit 2 and 26%, 18% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere due to gas-liquid partition and contribute to increase in the release to environment. A lot of evaluations for I-131 release have been performed so far by the MELCOR calculation or the SPEEDI reverse estimation. The SPEEDI reverse predicted significant release until 26 March, while no prediction in MELCOR after 17 March. The present study showed that iodine release from accumulated water may explain the release between 17 and 26 March. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks for several days after the core melt. (author)

  18. Overview of PNC studies of source term and sodium fires for FBRs

    It has been commonly understood that sodium plays an important role in the course of postulated accident progression in FBRs. Since 1970s, many efforts have been made at PNC-Japan to study sodium characteristics in various chemical and/or physical phenomena in FBR accidents. This paper reviews the research and development studies for source term and sodium fires performed at O-arai Engineering Center (OEC) of PNC

  19. Definition and evaluation of a dynamic source term module for use within RASTEP : A feasibility study

    Alfheim, Per

    2012-01-01

    RASTEP (RApid Source TErm Prediction) is a computerized tool for use in the fast diagnosis of accidents in nuclear power plants and analysis of the subsequent radiological source term. The tool is based on a Bayesian Belief Network that is used to determine the most likely plant state which in turn is associated with a pre-calculated source term from level 2 PSA. In its current design the source term predicting abilities of RASTEP are not flexible enough. Therefore, the purpose of this thesis...

  20. Estimation Of Source Term For Indian PHWRS (KAPS) As Part Of PSA Level-2 Study

    Source Term (ST) is generally known as the amount of the radio-nuclides(fission products along with activation and Actinides) that can be released from a nuclear power plant in an accident. The ST can be more accurately defined as the quantity, timing, composition, chemical and physical form of radio-nuclides. The amount of radio-nuclides is a fundamental parameter to estimate the consequences of an accident on individuals and environment. A quantitative estimation of the ST is of importance for assessing the effectiveness of safety design features and for the planning of post accident emergency measures in the public domain. The PSA Level-1 study for IPHWRs(KAPS) was completed in 2002 and an attempt was made to estimate the ST for different accident scenarios as part of PSA Level-2 study. The scope of this paper is limited to estimate the ST for Indian Pressurized Heavy Water Reactors (IPHWRs) in accident conditions. (authors)

  1. An appreciation of the events, models and data used for LMFBR radiological source term estimations

    In this report, the events, models and data currently available for analysis of accident source terms in liquid metal cooled fast neutron reactors are reviewed. The types of hypothetical accidents considered are the low probability, more extreme types of severe accident, involving significant degradation of the core and which may lead to the release of radionuclides. The base case reactor design considered is a commercial scale sodium pool reactor of the CDFR type. The feasibility of an integrated calculational approach to radionuclide transport and speciation (such as is used for LWR accident analysis) is explored. It is concluded that there is no fundamental obstacle, in terms of scientific data or understanding of the phenomena involved, to such an approach. However this must be regarded as a long-term goal because of the large amount of effort still required to advance development to a stage comparable with LWR studies. Particular aspects of LMFBR severe accident phenomenology which require attention are the behaviour of radionuclides during core disruptive accident bubble formation and evolution, and during the less rapid sequences of core melt under sodium. The basic requirement for improved thermal hydraulic modelling of core, coolant and structural materials, in these and other scenarios, is highlighted as fundamental to the accuracy and realism of source term estimations. The coupling of such modelling to that of radionuclide behaviour is seen as the key to future development in this area

  2. The source term experiments project deposition sample characterization: Interim report

    A series of four experiments aimed at characterizing the radiological source term associated with postulated severe light water reactor accidents has been conducted at Argonne National Laboratory's TREAT Facility. The experiments were designed to provide dta regarding the physicochemical properties, near the point of origin, of the biologically important volatile fission products released early in such accidents. The experimental vehicles were equipped to capture representative fission products released from fuel rods undergoing severe cladding degradation in a steam environment. Test conditions of pressure, fuel heatup rate, and steam flow were selected to simulate conditions predicted for hypothetical reactor accident sequences. One of the main components of the experimental vehicle's aerosol characterization system, common to all four tests, was a sample tree. It served to suspend coupons composed of a variety of materials, some typical of reactor structures, into the fission-product-laden steam flow. These coupons collected particulates and condensing vapors. Coupons frome ach of the four tests have been examined using scanning electron microscopy and associated energy dispersive x-ray analysis. The results of these initial sample examinations are presented. They are preceeded by a brief description of the test series and the experimental vehicle. Also included is a discussion of planned posttest examinations of other aerosol characterization system components and the test fuel as well as further examinations of the sample tree coupons. Results of the additional examinations thermal-hydraulic data, and interpretation of the information for each test will be included in future reports

  3. Contribution of steam explosions to the source term

    In the German Risk Study PWR (Phase A) a very important source term contribution was a steam explosion during a core meltdown accident. Detailed investigation has shown that four conditions must be present before a steam explosion resulting in extreme consequences can occur: (a) an extraordinarily large heat transfer area, (b) intimate contact between the melt and the water, (c) intensive heat transfer for a sufficiently long period of time, and (d) a sufficiently large mass of melt fragmenting in the water. If one of these conditions is not met, catastrophic consequences can be excluded. A synopsis proves that the first three conditions comply with the theoretical and experimental results for rather low pressures (order of magnitude: 2 bar) and small masses of melt (order of magnitude: 10 kg). However, these results are not necessarily valid for scaled-up conditions (e.g. tonnes of melt). For core-melt accidents under high pressure, spontaneous steam explosions are suppressed and there is no suitable trigger available for ignition. For low pressure conditions it has been proved that the amount of mass involved in the interaction (approximately 100 kg to 2 tonnes) is not sufficient to cause a steam explosion of catastrophic consequences. Therefore, PWR containments in the Federal Republic of Germany are not endangered by this phenomenon and steam explosions do not contribute to the source term. (author)

  4. Development of Reference Source Terms for EU-APR1400

    These source terms are developed for the typical U. S. NPP and do not reflect the design characteristics of EU-APR1400 (1,400 MWe PWR) which will be applied for the EUR certification in European countries. The process of developing the RST for EU-APR1400 is to undergo a similar process that NUREG-1465 had gone through when it came out with its proposed source terms. The purpose of this study is to develop the EU-APR1400 design-specific RST complied with the EUR. The Large LOCA is the reference equence used in the NUREG-1465 evaluation, whereas the EUAPR1400 risk-significant sequences are dominated by small LOCA and non-LOCA sequences. Moreover, when considering the EU-APR1400 has many design features to mitigate the consequences of severe accident phenomena, it is not surprising that the aspects of both release fractions and durations are distinctly different from NUREG-1465. This RST will be continuously updated to reflect to the design features of EU-APR1400, and then, be used as the reference for design purposes such as criteria satisfaction of radioactivity releases, equipment survivability, control room habitability for severe accident, and so on

  5. RBMK-1500 accident management for loss of long-term core cooling

    Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

  6. Source term estimation based on in-situ gamma spectrometry using a high purity germanium detector

    An alternative method to reconstruct the source term of a nuclear accident is proposed. The technique discussed here involves the use of in-situ gamma spectrometry. The validation of the applied methodology has been possible through the monitoring of routine releases of Ar-41 originating at a Belgian site from an air cooled graphite research reactor. This technique provides a quick nuclide specific decomposition of the source term and therefore will be have an enormous potential if implemented in nuclear emergency preparedness and radiological assessments of nuclear accidents during the early phase

  7. The long-term problems of contaminated land: Sources, impacts and countermeasures

    Baes, C.F. III

    1986-11-01

    This report examines the various sources of radiological land contamination; its extent; its impacts on man, agriculture, and the environment; countermeasures for mitigating exposures; radiological standards; alternatives for achieving land decontamination and cleanup; and possible alternatives for utilizing the land. The major potential sources of extensive long-term land contamination with radionuclides, in order of decreasing extent, are nuclear war, detonation of a single nuclear weapon (e.g., a terrorist act), serious reactor accidents, and nonfission nuclear weapons accidents that disperse the nuclear fuels (termed ''broken arrows'').

  8. The long-term problems of contaminated land: Sources, impacts and countermeasures

    This report examines the various sources of radiological land contamination; its extent; its impacts on man, agriculture, and the environment; countermeasures for mitigating exposures; radiological standards; alternatives for achieving land decontamination and cleanup; and possible alternatives for utilizing the land. The major potential sources of extensive long-term land contamination with radionuclides, in order of decreasing extent, are nuclear war, detonation of a single nuclear weapon (e.g., a terrorist act), serious reactor accidents, and nonfission nuclear weapons accidents that disperse the nuclear fuels (termed ''broken arrows'')

  9. STACE: Source Term Analyses for Containment Evaluations of transport casks

    The development of the Source Term Analyses for Containment Evaluations (STACE) methodology provides a unique means for estimating the probability of cladding breach within transport casks, quantifying the amount of radioactive material released into the cask interior, and calculating the releasable radionuclide concentrations and corresponding maximum permissible leakage rates. Following the guidance of ANSI N14.5, the STACE methodology provides a technically defensible means for estimating maximum permissible leakage rates. These containment criteria attempt to reflect the true radiological hazard by performing a detailed examination of the spent fuel, CRUD, and residual contamination contributions to the releasable source term. The evaluation of the spent fuel contribution to the source team has been modeled fairly accurately using the STACE methodology. The structural model predicts the cask drop load history, the mechanical response of the fuel assembly, and the probability of cladding breach. These data are then used to predict the amount of fission gas, volitile species, and fuel fines that are releasable from the cask. There are some areas where data are sparse or lacking in which experimental validation is planned. Finally, the ANSI N14.5 recommendation that 3% and 100% of the fuel rods fail during normal and hypothetical accident conditions of transport, respectively, has been show to be overly conservative by several degrees of magnitude for these example analyses. Furthermore, the maximum permissible leakage rates for this example assembly under normal and hypothetical accident conditions are significanly higher that the leaktight requirements. By relaxing the maximum permissible leakage rates, the source term methodology is expected to significantly improvecask economics and safety

  10. Light water reactor severe accident seminar. Seminar presentation manual

    The topics covered in this manual on LWR severe accidents were: Evolution of Source Term Definition and Analysis, Current Position on Severe Accident Phenomena, Current Position on Fission Product Behavior, Overview of Software Models Used in Severe Accident Analysis, Overview of Plant Specific Source Terms and Their Impact on Risk, Current Applications of Severe Accident Analysis, and Future plans

  11. Nuclear Powerplant Safety: Source Terms. Nuclear Energy.

    Department of Energy, Washington, DC. Nuclear Energy Office.

    There has been increased public interest in the potential effects of nuclear powerplant accidents since the Soviet reactor accident at Chernobyl. People have begun to look for more information about the amount of radioactivity that might be released into the environment as a result of such an accident. When this issue is discussed by people…

  12. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Andrej Prošek; Leon Cizelj

    2013-01-01

    Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pump...

  13. Nuclear fuel cycle facility accident analysis handbook

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  14. Emergency drinking water treatment during source water pollution accidents in China: origin analysis, framework and technologies.

    Zhang, Xiao-Jian; Chen, Chao; Lin, Peng-Fei; Hou, Ai-Xin; Niu, Zhang-Bin; Wang, Jun

    2011-01-01

    China has suffered frequent source water contamination accidents in the past decade, which has resulted in severe consequences to the water supply of millions of residents. The origins of typical cases of contamination are discussed in this paper as well as the emergency response to these accidents. In general, excessive pursuit of rapid industrialization and the unreasonable location of factories are responsible for the increasing frequency of accidental pollution events. Moreover, insufficient attention to environmental protection and rudimentary emergency response capability has exacerbated the consequences of such accidents. These environmental accidents triggered or accelerated the promulgation of stricter environmental protection policy and the shift from economic development mode to a more sustainable direction, which should be regarded as the turning point of environmental protection in China. To guarantee water security, China is trying to establish a rapid and effective emergency response framework, build up the capability of early accident detection, and develop efficient technologies to remove contaminants from water. PMID:21133359

  15. Remarks on methods of evaluation of aerosol sources related to PWR core meltdown accidents

    The paper tries to demonstrate the conceptional background of the KfK core melting program, which has been started in 1973, and which is scheduled to be terminated by 1986. The paper also summarizes the main findings of the SASCHA program, with the aid of which the enveloping fission product release from the primary system into the containment during a PWR core melt accident has been investigated. The fractions of release from the fuel determined in the experiment are undoubtedly in the range of 70% to 100% for the radiologically most important elements I, Cs, Te. The reduction in release from the primary circuit due to deposition is 50% at the maximum. A considerable portion resuspended must be deducted from that value. The retention of iodine and aerosol particles in the safety containment amounts to several orders of magnitude (up to 5). Likewise, the decrease in the population dose by spread and dilution in the environment and due to other parameters attains several orders of magnitude (up to 7). Consequently, particle retention by a factor of 2 or 3 in the primary circuit is negligible. - Our present knowledge is completely satisfactory for analyzing the so-called source term in core melt accidents. The wish to develop more detailed codes related to core degradation and to activity release from the primary circuit has many understandable causes. However, there is no single technical reason in favor of spending much money in order to materialize this wish. (orig./HP)

  16. Subsurface Shielding Source Term Specification Calculation

    The purpose of this calculation is to establish appropriate and defensible waste-package radiation source terms for use in repository subsurface shielding design. This calculation supports the shielding design for the waste emplacement and retrieval system, and subsurface facility system. The objective is to identify the limiting waste package and specify its associated source terms including source strengths and energy spectra. Consistent with the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M and O 2001, p. 15), the scope of work includes the following: (1) Review source terms generated by the Waste Package Department (WPD) for various waste forms and waste package types, and compile them for shielding-specific applications. (2) Determine acceptable waste package specific source terms for use in subsurface shielding design, using a reasonable and defensible methodology that is not unduly conservative. This calculation is associated with the engineering and design activity for the waste emplacement and retrieval system, and subsurface facility system. The technical work plan for this calculation is provided in CRWMS M and O 2001. Development and performance of this calculation conforms to the procedure, AP-3.12Q, Calculations

  17. Considerations about the implementation of alternative source terms: nuclear safety and plant operational benefits

    In this paper, several aspects are discussed about the implementation of an alternative source term for the analysis of the radiological consequences of design basis accidents in nuclear power plants. First, the rationale for implementation of an alternative source term is discussed. Then, the topics studied start by considering the current methodology and regulation applied to determine the original source term. Next, to determine a different source term, the basis of a new methodology is discussed, as, for example the elimination of excessive conservative assumptions. As a consequence of the adoption of an alternative source term, operational benefits are expected from relaxation of regulatory requirements established in the plant technical specifications. Other key issues considered in this work are the use of engineered safety features to minimize the iodine release during an accident, and technical requirements regarding the safe operation of the emergency filtering system for the main control room, in order to protect the reactor operation personnel. Finally, a discussion is presented about the impact on risk assessment, when using an alternative source term, and remarking that the adoption of a new source term by itself do not have and impact on plant risk, but it does have an effect on radiological consequences. Nevertheless, a detailed review of technical specification changes that could induce some risk should be considered. As conclusions of this work, recommendations are presented for the licensing process of an alternative source term. (Author)

  18. Management for the prevention of accidents from disused sealed radioactive sources

    The objective of this report is to provide advice to sealed radiation source (SRS) users, radioactive waste operators, and other concerned public sectors on the measures to be taken to reduce the risk of accidents associated with disused or spent SRS. The report also explains policies as well as technical and administrative procedures to minimize the risk of accidents and to mitigate the consequences should an accident occur. The report emphasizes areas of high risk in handling disused or spent SRS in any form and condition to help to save health, life and financial resources

  19. New source terms: what do they tell us about engineered safety feature performance

    The accident behavior models which are the basis of engineered safety feature design are generally simple, non-mechanistic and concentrated on volatile radioiodine. Now data from source term studies show that models should be more mechanistic and look at other species than volatile iodine. A complete reevaluation of engineered safety features is needed

  20. Sensitivities to source-term parameters of emergency planning zone boundaries for waste management facilities

    This paper reviews the key parameters comprising airborne radiological and chemical release source terms, discusses the ranges over which values of these parameters occur for plausible but severe waste management facility accidents, and relates the concomitant sensitivities of emergency planning zone boundaries predicted on calculated distances to early severe health effects

  1. Licensing design basis source term update for the Evolutionary Advanced Light Water Reactor

    The purpose of this report is to document the technical basis for a licensing source term update for the Evolutionary Advanced Light Water Reactor (ALWR) which will make the source term more physically realistic. While TID [Technical Information Document] 14844 and related regulatory guidance have served the industry well, much has been learned about source term over the last 30 years, and the ALWR Requirements Document provides an opportunity to incorporate this experience by updating the licensing source term. Further, the source term update will provide an improved basis for evolutionary ALWR accident mitigation design. Results of this work indicate that the fission product release magnitude to containment is slightly less than TID 14844 for noble gas, iodine, and semi and low volatiles, but somewhat higher for cesium and tellurium. Release timing is delayed by one hour or more after the accident initiation. The chemical form of iodine is largely aerosol with significantly less organic iodine compared to regulatory guidance which specifies mostly elemental and a relatively large fraction of organic. Containment spray aerosol removal rate was determined to be significantly higher than specified in regulatory guidance. Finally, BWR suppression pool decontamination factor was determined to be less effective than allowed by regulatory guidance early in the accident (due to the delayed release noted above) and more effective than that allowed by regulatory guidance later in the accident. It is recognized by the ALWR program that the source term update could be taken further in the direction of a physically-based source term. 47 refs., 4 figs., 11 tabs

  2. Assessment of the efficiency of short term countermeasures following a severe accident on a PWR

    In case of a severe nuclear accident at a PWR plant, countermeasures will be initiated in the short term by authorities to reduce the consequences of the atmospheric radioactive releases on the neighbouring population. Various factors influence the level of protection afforded by countermeasures. For instance, a too late intervention would lead to a Jack of efficiency in terms of dose reduction if the actual evolution of the accident is not considered. Thus, implementation of countermeasures should be optimized. In general, the projected doses (those without countermeasure) are compared with those expected when a particular countermeasure or strategy is implemented. In this paper, an in-depth analysis associates the kinetics of the release with the corresponding evolution of the dosimetric efficiency of countermeasures. This is done at different times in the short term of the accident and for various distances from the plant. Results are presented for different strategies initiated at various times. This work gives useful information for the early management of a major nuclear accident. (authors)

  3. Radiation source term reduction in BWR plants

    This series of slides presents: the collective radiation exposures at US and European BWRs; the European experience with source term reduction measures (normal water chemistry - NWC): zinc addition, stellite replacement, full system decontamination; the effects of evolving water chemistries/US experience. The conclusions are summarized as follows: worldwide reduction of collective radiation exposures at BWRs by following the ALARA principle; zinc addition proven option for source term reduction for NWC and hydrogen water chemistry (HWC) plants; reducing feedwater iron has been proven to reduce dose rates - as operational observations in the US indicate; optimized feedwater iron is very important for fuel performance under all modes of water chemistry (HWC, Zn, and noble metal chemical addition (NMCA)); minimize 59Co sources/stellite, follow the ALARA principle; full system decontamination (FSD) plus zinc injection is an attractive option for reducing reactor coolant system (RCS) dose rates of mature BWR plants

  4. The rehabilitation strategies in agriculture in the long term after the Chernobyl NPP accident

    The experience gained in the aftermath of the severe radiation accidents shows that in the case of large-scaled radionuclide contamination the limitation of internal radiation doses to people by means of restoration of agricultural lands is more realistic than reduction of levels of external irradiation. Therefore, the problems connected with the optimal restoration strategies of agricultural land subjected to radioactive contamination after the Chernobyl accident are of crucial importance. The justification of the approach for the estimation of the effectiveness of countermeasure strategies in the long term after the Chernobyl accident, based on the classification of farms by contamination density and risk of the exceeding of radiological standards, restricting the use of agricultural products, is presented. For each class of the farms the ranking of rehabilitation options and the time periods when their application would be of importance are given. Comparative analysis of the rehabilitation strategies, which are different in their effectiveness and cost, is provided. (author)

  5. Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1

    Pappas, Charalampos, E-mail: chpappas@ipta.demokritos.gr; Ikonomopoulos, Andreas; Sfetsos, Athanasios; Andronopoulos, Spyros; Varvayanni, Melpomeni; Catsaros, Nicolas

    2014-07-01

    Highlights: • Source term derivation of postulated accident sequences in a research reactor. • Various containment ventilation scenarios considered for source term calculations. • Source term parametric analysis performed in case of lack of ventilation. • JRODOS employed for dose calculations under eighteen modeled scenarios. • Estimation of radiological consequences during typical and adverse weather scenarios. - Abstract: The estimated source term, dose results and radiological consequences of selected accident sequences in the Greek Research Reactor – 1 are presented and discussed. A systematic approach has been adopted to perform the necessary calculations in accordance with the latest computational developments and IAEA recommendations. Loss-of-coolant, reactivity insertion and fuel channel blockage accident sequences have been selected to derive the associated source terms under three distinct containment ventilation scenarios. Core damage has been conservatively assessed for each accident sequence while the ventilation has been assumed to function within the efficiency limits defined at the Safety Analysis Report. In case of lack of ventilation a parametric analysis is also performed to examine the dependency of the source term on the containment leakage rate. A typical as well as an adverse meteorological scenario have been defined in the JRODOS computational platform in order to predict the effective, lung and thyroid doses within a region defined by a 15 km radius downwind from the reactor building. The radiological consequences of the eighteen scenarios associated with the accident sequences are presented and discussed.

  6. Derivation of the source term, dose results and associated radiological consequences for the Greek Research Reactor – 1

    Highlights: • Source term derivation of postulated accident sequences in a research reactor. • Various containment ventilation scenarios considered for source term calculations. • Source term parametric analysis performed in case of lack of ventilation. • JRODOS employed for dose calculations under eighteen modeled scenarios. • Estimation of radiological consequences during typical and adverse weather scenarios. - Abstract: The estimated source term, dose results and radiological consequences of selected accident sequences in the Greek Research Reactor – 1 are presented and discussed. A systematic approach has been adopted to perform the necessary calculations in accordance with the latest computational developments and IAEA recommendations. Loss-of-coolant, reactivity insertion and fuel channel blockage accident sequences have been selected to derive the associated source terms under three distinct containment ventilation scenarios. Core damage has been conservatively assessed for each accident sequence while the ventilation has been assumed to function within the efficiency limits defined at the Safety Analysis Report. In case of lack of ventilation a parametric analysis is also performed to examine the dependency of the source term on the containment leakage rate. A typical as well as an adverse meteorological scenario have been defined in the JRODOS computational platform in order to predict the effective, lung and thyroid doses within a region defined by a 15 km radius downwind from the reactor building. The radiological consequences of the eighteen scenarios associated with the accident sequences are presented and discussed

  7. Station blackout analysis of nuclear power plant using source term code package

    Continuous efforts to ensure the safety of nuclear installations in Slovenia have led to comprehensive analysis of Levels II and III of hypothetic station blackout accident modelled using the tools at our disposal. This paper represents the thermal hydraulic and radionuclide transport part of the overall effort. MARCH3 and VANESA modules of Source Term Code Package were used to analyze four different scenario depending on different reactor coolant pump leak rate (125 gpm and 400 gpm, respectively) and containment design pressure (i.e. 0.309 Mpa and 0.785 Mpa). The final aim of the project was to prepare input into the Level III analyses of the accident. The accident starts by loss of off-site power combined with loss of diesel generators. The turbine driven auxiliary feedwater pump operates additional two hours after the inception of the accident. The results are given in form of graphs displaying reactor coolant system and containment parameters. (author)

  8. TREAT light water reactor source term experiments program

    Pre-test calculations indicate that, for the STEP-1 (Source Term Experimental Program) test, cladding temperatures in excess of 42000F can be reached on a heatup transient similar to that of the AD accident sequence in a 20-min test duration. This is well above the Zircaloy melting point of approx. 33500F and should provide a degree of cladding disruption sufficient to allow a singificant release of products from the fuel into the flowing steam. The same temperature range can be reached in a 60-min-duration run to simulate the TQUW sequence for the STEP-2 test. The complete paper will present initial experimental results from these two tests and perhaps from the two TMLB' simulations run without and with control rod material in STEP-3 and STEP-4, respectively

  9. Source-term evaluations from recent core-melt experiments

    Predicted consequences of hypothetical severe reactor accidents resulting in core meltdown appear to be too conservatively projected because of the simplistic concepts often assumed for the intricate and highly variable phenomena involved. Recent demonstration work on a modest scale (1-kg) has already revealed significant variations in the mode and temperature for clad failure, in the rates of formation of zirconium alloys, in the nature of the UO2-ZrO2 eutectic mixtures, and in aerosol generation rates. The current series of core-melt demonstration experiments (at the 10-kg scale) seem to confirm that an increase in size of the meltdown mass will lead to an even further reduction in the amount of vaporized components. Source terms that are based on older release evaluations could be up to an order of magnitude too large. 6 refs., 6 figs., 2 tabs

  10. The Phebus Fission Product and Source Term International Programmes

    The international Phebus FP programme, initiated in 1988 is one of the major research programmes on light water reactors severe accidents. After a short description of the facility and of the test matrix, the main outcomes and results of the first four integral tests are provided and analysed. Several results were unexpected and some are of importance for safety analyses, particularly concerning fuel degradation, cladding oxidation, chemical form of some fission products, especially iodine, effect of control rod materials on degradation and chemistry, iodine behaviour in the containment. Prediction capabilities of calculation tools have largely been improved as a result of this research effort. However, significant uncertainties remain for a number of phenomena, requiring detailed physical analysis and implementation of improved models in codes, sustained by a number of separate-effect experiments. This is the subject of the new Source Term programme for a better understanding of the phenomenology on important safety issues, in accordance with priorities defined in the EURSAFE project of the 5th European framework programme aiming at reducing the uncertainties on Source Term analyses. It covers iodine chemistry, impact of boron carbide control rods degradation and oxidation, air ingress situations and fission product release from fuel. Regarding the interpretation of Phebus, an international co-operation has been established since over ten years, particularly helpful for the improvement and common understanding of severe accident phenomena. Few months ago, the Phebus community was happy to welcome representatives of a large number of organisations from the following new European countries: the Czech republic, Hungary, Lithuania, Slovakia, Slovenia and also from Bulgaria and Romania. (author)

  11. Verification test calculations for the Source Term Code Package

    The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs

  12. Source term evaluations from recent core-melt experiments

    Predicted consequences of hypothetical severe reactor accidents resulting in core meltdown appear to be too conservatively projected because of the simplistic concepts often assumed for the intricate and highly variable phenomena involved. Recent demonstration work on a modest scale (1 kg) has already revealed significant variations in the mode and temperature for clad failure, in the rates of formation of zirconium alloys, in the nature of UO2-ZrO2 eutectic mixtures and in aerosol generation rates. In PWRs, these aspects of core degradation are dependent upon the extent and nature of the dispersion of the control rod silver alloy and the interaction of both silver and molten stainless steel with the Zircaloy-processes which are overlooked. In fast-melt accident sequences, the dominant aerosol species generated in the early stages of core meltdown (before primary vessel failure) now appear to be cadmium, manganese, silver and the volatile fission products rather than structural or fuel materials, uranium oxide or zirconium. The current series of core-melt demonstration experiments (at the 10 kg scale) seem to confirm that an increase in size of the meltdown mass will lead to an even further reduction in the amount of vaporized components. Source terms that are based on older release evaluations could be up to an order of magnitude too large. (author)

  13. BWR Source Term Generation and Evaluation

    This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has the document identifier BBAC00000-01717-0210-000061. The purpose of this revision is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version (Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one million years after the SNF is discharged from the reactors. This calculation includes an examination of several ways to represent BWR assemblies and operating conditions in SAS2H in order to quantify the effects these representations may have on source terms. These source terms provide information characterizing the neutron and gamma spectra in particles per second, the decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a range of burnups and enrichments (see Table 2) that are representative of the waste stream and stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups used for the computer runs of the previous revision were actually about 1.7% less than the stated, or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before using any source terms from this calculation. The source term due to the activation of corrosion products deposited on the surfaces of the assembly from the coolant is also calculated. The results of this calculation support many areas of the Monitored Geologic Repository (MGR), which include thermal evaluation, radiation dose determination, radiological safety analyses, surface and subsurface facility designs, and total system performance assessment. This includes MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements of the Quality

  14. Using Bayesian Belief Network (BBN) modelling for Rapid Source Term Prediction. RASTEP Phase 1

    Knochenhauer, M.; Swaling, V.H.; Alfheim, P. [Scandpower AB, Sundbyberg (Sweden)

    2012-09-15

    The project is connected to the development of RASTEP, a computerized source term prediction tool aimed at providing a basis for improving off-site emergency management. RASTEP uses Bayesian belief networks (BBN) to model severe accident progression in a nuclear power plant in combination with pre-calculated source terms (i.e., amount, timing, and pathway of released radio-nuclides). The output is a set of possible source terms with associated probabilities. In the NKS project, a number of complex issues associated with the integration of probabilistic and deterministic analyses are addressed. This includes issues related to the method for estimating source terms, signal validation, and sensitivity analysis. One major task within Phase 1 of the project addressed the problem of how to make the source term module flexible enough to give reliable and valid output throughout the accident scenario. Of the alternatives evaluated, it is recommended that RASTEP is connected to a fast running source term prediction code, e.g., MARS, with a possibility of updating source terms based on real-time observations. (Author)

  15. Source term modifications by re-entrainment aerosols from boiling sump water

    Entrainment of droplets from the containment pool by gas flow under boiling or bubbling conditions is of importance in the source term evaluation of a loss-of-coolant accident with late containment failure caused by overpressurization. Most of the particulate fraction of the fission products released from the core melt is deposited in the containment during the enclosure time and is assumed to be transported into the containment sump, which has been heated by decay heat since the core melt/concrete interaction period. A sensitivity study of the aerosol source term is performed by varying the two key factors of droplet entrainment, namely the particle size and entrainment fraction. Using a suitable set of parameters the dominant source term contribution of caesium, for example in the above accident is evaluated at 3x10-4 fractions of the core inventory. (author)

  16. SUBURFACE SHIELDING-SPECIFIC SOURCE TERM EVALUATION

    The purpose of this work is to provide supporting calculations for determination of the radiation source terms specific to subsurface shielding design and analysis. These calculations are not intended to provide the absolute values of the source terms, which are under the charter of the Waste Package Operations (WPO) Group. Rather, the calculations focus on evaluation of the various combinations of fuel enrichment, burnup and cooling time for a given decay heat output, consistent with the waste package (WP) thermal design basis. The objective is to determine the worst-case combination of the fuel characteristics (enrichment, burnup and cooling time) which would give the maximum radiation fields for subsurface shielding considerations. The calculations are limited to PWR fuel only, since the WP design is currently evolving with thinner walls and a reduced heat load as compared to the viability assessment (VA) reference design. The results for PWR fuel will provide a comparable indication of the trend for BWR fuel, as their characteristics are similar. The source term development for defense high-level waste and other spent nuclear fuel (SNF) is the responsibility of the WPO Group, and therefore, is not included this work. This work includes the following items responsive to the stated purpose and objective: (1) Determine the possible fuel parameters (initial enrichment, burnup and cooling time), that give the same decay heat value as specified for the waste package thermal design; (2) Obtain the neutron and gamma source terms for the various combinations of the fuel parameters for use in radiation field calculations; and (3) Calculate radiation fields on the surfaces of the waste package and its transporter to quantify the effects of the fuel parameters with the same decay heat value for use in identifying the worst-case combination of the fuel parameters

  17. ESTER - a European source term evaluation system

    The Commission of the European Communities (CEC) sponsors considerable model development and validation in the area of Light Water Reactor (LWR) source term, and naturally wishes to see the results used as widely as possible. It also has a role in fostering collaboration between European teams involved in source term analysis, for which purpose Phebus-Fission Product (FP) is acting as a focal point. To further both aims the Joint Research Centre (JRC) decided in 1989 to sponsor the development of the best-estimate code ESTER, which is both a software environment and a set of coupled source term modules which when completed should offer potentialities not currently available within Europe. This paper describes first the overall architecture of ESTER, then the component parts: the tools and services, the user interface, and the modules which perform the physics and chemistry calculations, emphasizing the design choices which have been made. The quality assurance system for the whole system is also reviewed. Contributions from the model developers, both underway, and expected, are then surveyed in the context of the overall development of ESTER, and the planning of the creation and extension of ESTER is given. The paper closes with some proposals for sharing ESTER within Europe and for ensuring its maintenance and continued rational development. (Author)

  18. Hazardous constituent source term. Revision 2

    1994-11-17

    The Department of Energy (DOE) has several facilities that either generate and/or store transuranic (TRU)-waste from weapons program research and production. Much of this waste also contains hazardous waste constituents as regulated under Subtitle C of the Resource Conservation and Recovery Act (RCRA). Toxicity characteristic metals in the waste principally include lead, occurring in leaded rubber gloves and shielding. Other RCRA metals may occur as contaminants in pyrochemical salt, soil, debris, and sludge and solidified liquids, as well as in equipment resulting from decontamination and decommissioning activities. Volatile organic compounds (VOCS) contaminate many waste forms as a residue adsorbed on surfaces or occur in sludge and solidified liquids. Due to the presence of these hazardous constituents, applicable disposal regulations include land disposal restrictions established by Hazardous and Solid Waste Amendments (HSWA). The DOE plans to dispose of TRU-mixed waste from the weapons program in the Waste Isolation Pilot Plant (WIPP) by demonstrating no-migration of hazardous constituents. This paper documents the current technical basis for methodologies proposed to develop a post-closure RCRA hazardous constituent source term. For the purposes of demonstrating no-migration, the hazardous constituent source term is defined as the quantities of hazardous constituents that are available for transport after repository closure. Development of the source term is only one of several activities that will be involved in the no-migration demonstration. The demonstration will also include uncertainty and sensitivity analyses of contaminant transport.

  19. Spent fuel assembly source term parameters

    Containment of cask contents by a transport cask is a function of the cask body, one or more closure lids, and various bolting hardware, and seals associated with the cavity closure and other containment penetrations. In addition, characteristics of cask contents that impede the ability of radionuclides to move from an origin to the external environment also provide containment. In essence, multiple release barriers exist in series in transport casks, and the magnitude of the releasable activity in the cask is considerably lower than the total activity of its contents. A source term approach accounts for the magnitude of the releasable activity available in the cask by assessing the degree of barrier resistance to release provided by material characteristics and inherent barriers that impede the release of radioactive contents. Standardized methodologies for defining the spent-fuel transport packages with specified regulations have recently been developed. An essential part of applying the source term methodology involves characterizing the response of the spent fuel under regulatory conditions of transport. Thermal and structural models of the cask and fuel are analyzed and used to predict fuel rod failure probabilities. Input to these analyses and failure evaluations cover a wide range of geometrical and material properties. An important issue in the development of these models is the sensitivity of the radioactive source term generated during transport to individual parameters such as temperature and fluence level. This paper provides a summary of sensitivity analyses concentrating on the structural response and failure predictions of the spent fuel assemblies

  20. Long-term decline of 137Cs concentration in honey in the second decade after the Chernobyl accident

    In the years 2001-2004 the 137Cs activity was investigated in a total of 336 samples of different varieties of honey harvested in the Liguria Region of Northern Italy. Our purpose was to define (a) residual radioactive contamination following the Chernobyl accident and 137Cs long-term decline, (b) correlation between 137Cs activity and different honey varieties, and (c) correlation between 137Cs activity and the prevailing geomorphological configuration in the collection areas. The mean 137Cs specific activity was 4.33 ± 5.04 S.D. Bq/kg. Chestnut honey showed higher levels of radioactive contamination, which were ascribed to the extensive, superficial and deep, root apparatus of the tree. Honey samples from acidic argillite soils, which withhold radionuclides after deposition and slowly release them to plants, also showed higher 137Cs activity. Long-term decline was calculated at 456 days, a value lower than those published from different food sources in the years following the accident. The rate of long-term decline decreases with time

  1. Radiological consequence analyses under severe accident conditions for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    This paper discusses salient aspects of methodology, assumptions, modeling of various features related to radiation exposure, and health consequences from source terms resulting from two conservatively scoped severe accident scenarios. Radiological consequences for a site-suitability scenario based on 10 CFR 100 guidelines also are presented. Consequences arising from severe accidents involving steaming pools and core-concrete interaction (CCI) events combined with several different containment configurations are presented. Results are presented in the form of mean cumulative values for prompt and latent cancer fatality estimates and related cumulative, complementary distribution functions as a function of distance from the reactor site. It is shown that the reactor-site-suitability risk goals are met by a large margin and that overall risk is dominated by early containment failure combined with CCI events

  2. IMPACTS OF SOURCE TERM HETEROGENEITIES ON WATER PATHWAY DOSE.

    SULLIVAN, T.; GUSKOV, A.; POSKAS, P.; RUPERTI, N.; HANUSIK, V.; ET AL.

    2004-09-15

    specific activities and small physical sizes and for which a solution has to be found in term of long-term disposal. Together with their casing and packaging, they are one form of heterogeneous waste; many other forms of waste with heterogeneous properties exist. They may arise in very small quantities and with very specific characteristics in the case of small producers, or in larger streams with standard characteristics in others. This wide variety of waste induces three main different levels of waste heterogeneity: (1) hot spot (e.g. disused sealed sources); (2) large item inside a package (e.g. metal components); and (3) very large items to be disposed of directly in the disposal unit (e.g. irradiated pipes, vessels). Safety assessments generally assume a certain level of waste homogeneity in most of the existing or proposed disposal facilities. There is a need to evaluate the appropriateness of such an assumption and the influence on the results of safety assessment. This need is especially acute in the case of sealed sources. There are many cases where are storage conditions are poor, or there is improper management leading to a radiological accident, some with significant or detrimental impacts. Disposal in a near surface disposal facility has been used in the past for some disused sealed sources. This option is currently in use for others sealed sources, or is being studied for the rest of them. The regulatory framework differs greatly between countries. In some countries, large quantities of disused sealed sources have been disposed of without any restriction, in others their disposal is forbidden by law. In any case, evaluation of the acceptability of disposal of disused sealed sources in near surface disposal facility is of utmost importance.

  3. Using Bayesian Belief Network (BBN) modelling for rapid source term prediction. Final report

    Knochenhauer, M.; Swaling, V.H.; Dedda, F.D.; Hansson, F.; Sjoekvist, S.; Sunnegaerd, K. [Lloyd' s Register Consulting AB, Sundbyberg (Sweden)

    2013-10-15

    The project presented in this report deals with a number of complex issues related to the development of a tool for rapid source term prediction (RASTEP), based on a plant model represented as a Bayesian belief network (BBN) and a source term module which is used for assigning relevant source terms to BBN end states. Thus, RASTEP uses a BBN to model severe accident progression in a nuclear power plant in combination with pre-calculated source terms (i.e., amount, composition, timing, and release path of released radio-nuclides). The output is a set of possible source terms with associated probabilities. One major issue has been associated with the integration of probabilistic and deterministic analyses are addressed, dealing with the challenge of making the source term determination flexible enough to give reliable and valid output throughout the accident scenario. The potential for connecting RASTEP to a fast running source term prediction code has been explored, as well as alternative ways of improving the deterministic connections of the tool. As part of the investigation, a comparison of two deterministic severe accident analysis codes has been performed. A second important task has been to develop a general method where experts' beliefs can be included in a systematic way when defining the conditional probability tables (CPTs) in the BBN. The proposed method includes expert judgement in a systematic way when defining the CPTs of a BBN. Using this iterative method results in a reliable BBN even though expert judgements, with their associated uncertainties, have been used. It also simplifies verification and validation of the considerable amounts of quantitative data included in a BBN. (Author)

  4. Using Bayesian Belief Network (BBN) modelling for rapid source term prediction. Final report

    The project presented in this report deals with a number of complex issues related to the development of a tool for rapid source term prediction (RASTEP), based on a plant model represented as a Bayesian belief network (BBN) and a source term module which is used for assigning relevant source terms to BBN end states. Thus, RASTEP uses a BBN to model severe accident progression in a nuclear power plant in combination with pre-calculated source terms (i.e., amount, composition, timing, and release path of released radio-nuclides). The output is a set of possible source terms with associated probabilities. One major issue has been associated with the integration of probabilistic and deterministic analyses are addressed, dealing with the challenge of making the source term determination flexible enough to give reliable and valid output throughout the accident scenario. The potential for connecting RASTEP to a fast running source term prediction code has been explored, as well as alternative ways of improving the deterministic connections of the tool. As part of the investigation, a comparison of two deterministic severe accident analysis codes has been performed. A second important task has been to develop a general method where experts' beliefs can be included in a systematic way when defining the conditional probability tables (CPTs) in the BBN. The proposed method includes expert judgement in a systematic way when defining the CPTs of a BBN. Using this iterative method results in a reliable BBN even though expert judgements, with their associated uncertainties, have been used. It also simplifies verification and validation of the considerable amounts of quantitative data included in a BBN. (Author)

  5. Health physics evaluation of an accident involving acute overexposure to a radiography source

    An accident, involving the loss of an iridium-192 radiographic source and the subsequent serious overexposure of a third party, is described. Health physics aspects, particularly dosimetrical aspects are addressed and compared with results obtained by means of chromosome aberration dosimetry. Details are provided on the medical observations and treatment of the patient

  6. Implications of source term research for ex-plant consequence modeling

    Current fission product source term research has important implications for several areas of ex-plant consequence modeling, including predictions of (a) the public risk of early fatality, early injury, and latent cancer fatality; (b) individual risk of early and latent cancer fatality and comparison with safety goals; (c) deposition of radioactive materials and subsequent contamination of land; (d) provision of technical input to considerations of emergency response and planning. The sensitivity of the above to source term characteristics (e.g., source term magnitudes, timing of the release, and duration of the release) will be investigated. Significant issues will be discussed in terms of current source term research; e.g., whether such research yet demonstrates that early fatalities can be ruled out after a reactor accident

  7. Important severe accident research issues after Fukushima accident

    After the Fukushima accident several investigation committees issued reports with lessons learned from the accident in Japan. Among those lessons, several recommendations have been made on severe accident research. Similar to the EURSAFE efforts under EU Program, review of specific severe accident research items was started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation has been started since the Fukushima accident. Additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality have been covered. The review results are categorized in ten major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, direct containment heating, recriticality and instrumentation in severe accident conditions. Based on these activities and also author's personal view, the present paper describes the perspective of important severe accident research issues after Fukushima accident. Those are specifically investigation of damaged core and components, advanced severe accident analysis capabilities and associated experimental investigations, development of reliable passive cooling system for core/containment, analysis of hydrogen behavior and investigation of hydrogen measures, enhancement of removal function of radioactive materials of containment venting, advanced instrumentation for the diagnosis of severe accident and assessment of advanced containment design which excludes long-term evacuation in any severe accident situations. (author)

  8. Analysis of Hydrogen Source Term and Effectiveness of Hydrogen Control in Thousand Megawatt PWR Severe Accident%百万千瓦级压水堆严重事故下氢气源项及氢气空制有效性分析

    邹杰; 佟立丽; 曹学武; 顾健; 薛峻峰; 江宇; 郝禄禄; 仇苏辰; 刘力

    2013-01-01

    针对百万千瓦级压水堆核电厂大型干式安全壳在严重事故情况下的氢气风险控制,建立了一体化事故分析模型,分别对大破口失水事故(LB-LOCA)、中破口失水事故(MB-LOCA)、小破口失水事故(SB-LOCA)、全厂断电事故(SBO)、蒸汽发生器(SG)传热管破裂事故(SGTR)以及主蒸汽管道破裂事故(MSLB)进行事故进程计算以及氢气源项分析.相对于其他事故序列,LB-LOCA下堆芯快速熔化,锆-水反应产生氢气的速率快,可以作为安全壳内氢气风险控制有效性分析的代表性事故序列.分析表明,严重事故情况下在安全壳中安装一定数量的非能动氢气复合器(PARs)能够有效去除安全壳中的氢气,消除氢气燃烧或爆炸的风险,保持安全壳的完整性.%The integrated severe accident analysis model of 100 MW PWR NPP is built to analyze the hydrogen risk under severe accidents.Large break loss of coolant accident (LB-LOCA),medium break loss of coolant accident (MB-LOCA),small break loss of coolant accident (SB-LOCA),station blackout (SBO),steam generator tube rupture (SGTR) and main steam line break (MSLB) are chosen as typical severe accident sequences to analyze the hydrogen source.Considering the hydrogen quantity of 100% zirconium water reaction,the LB-LOCA is selected as a representative sequence to evaluate the hydrogen mitigation system.The results show that a certain number of PARs could remove hydrogen and oxygen effectively,and protect the containment integrity against hydrogen deflagration or detonation.

  9. Transport accident frequency data, their sources and their application in risk assessment

    Base transport accident frequency data and sources of these data are presented. Both generic information and rates specific to particular routes or packages are included. Strong packages, such as those containing significant quantities of radioactive materials, will survive most of the accidents represented by these base frequencies without a containment breach. The association of severity probability distributions with a base frequency, and package and contents response, leading to the quantification of release frequency and magnitude, are often more important in risk assessment than the base frequency itself. This paper therefore also includes brief comments on techniques adopted to utilize the base frequencies. This paper reports an accident frequency data survey undertaken at the end of 1986. It has not been updated to take account of work published between January 1987 and the Report publication date. (author)

  10. Importance of core/concrete aerosol production and some containment heat sources to the source term

    Production of aerosols by core/concrete interaction in a large break PWR severe accident is discussed, and both vaporization and mechanical production processes are examined. In the case of the former, equilibrium chemical thermodynamic studies are used to decide which chemical species should be considered, recognizing the uncertainty in the likely configuration of the core/concrete melt. Lanthanide release is found to be particularly sensitive to this configuration. It is found that kinetic effects are not important in preventing the attainment of chemical equilibrium in the gas bubbling through the melt. At early times aerosol production by bubble bursting at the melt surface is found to be less important than that due to vaporization, except for those materials released in small quantities, e.g. Mo. The bubble bursting mechanism becomes relatively more important at later times. Calculations for a large modern PWR show that environmental release from the core/concrete aerosol is likely to be of comparable or greater importance (in terms of released decay heat) than that from the in-vessel core-melt aerosol for all but very early containment failure or failure to isolate, neglecting attenuation of the core/concrete aerosol during its flow from the cavity to the main containment volume. The importance of performing linked thermal-hydraulic and aerosol physics calculations is highlighted by the blowdown aerosol in a large break accident. Treatment of the decay heat arising from the aerosol material released to the containment is discussed. It is shown that it is very important to consider this heat source in containment pressure calculations, but it was not found to be important to treat its spatial dependence accurately in the large break accident considered here. Some scoping calculations for material resuspension on containment overpressure failure, due to a hydrogen burn, are presented

  11. Reassessment of the technical bases for estimating source terms. Draft report for comment

    NUREG-0956 describes the NRC staff and contractor efforts to reassess and update the agency's analytical procedures for estimating accident source terms for nuclear power plants. The effort included development of a new source term analytical procedure - a set of computer codes - that is intended to replace the methodology of the Reactor Safety Study (WASH-1400) and to be used in reassessing the use of TID-14844 assumptions (10 CFR 100). NUREG-0956 describes the development of these codes, the demonstration of the codes to calculate source terms for specific cases, the peer review of this work, some perspectives on the overall impact of new source terms on plant risks, the plans for related research projects, and the conclusions and recommendations resulting from the effort

  12. Source term evaluation for combustion modeling

    Sussman, Myles A.

    1993-01-01

    A modification is developed for application to the source terms used in combustion modeling. The modification accounts for the error of the finite difference scheme in regions where chain-branching chemical reactions produce exponential growth of species densities. The modification is first applied to a one-dimensional scalar model problem. It is then generalized to multiple chemical species, and used in quasi-one-dimensional computations of shock-induced combustion in a channel. Grid refinement studies demonstrate the improved accuracy of the method using this modification. The algorithm is applied in two spatial dimensions and used in simulations of steady and unsteady shock-induced combustion. Comparisons with ballistic range experiments give confidence in the numerical technique and the 9-species hydrogen-air chemistry model.

  13. TRIGA MARK-II source term

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel

  14. TRIGA MARK-II source term

    Full-text: ORIGEN 2.2 are employed to obtain data regarding g source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel. (author)

  15. TRIGA MARK-II source term

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my; Rawi, M. Z. M. Rawi, E-mail: mark-dennis@nuclearmalaysia.gov.my; Abu, M. P., E-mail: mark-dennis@nuclearmalaysia.gov.my [Bahagian Teknologi Reaktor, Agensi Nuklear Malaysia, 43000 Kajang (Malaysia)

    2014-02-12

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  16. TRIGA MARK-II source term

    Usang, M. D.; Hamzah, N. S.; J. B., Abi M.; M. Z., M. Rawi; Abu, M. P.

    2014-02-01

    ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences of results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.

  17. Use of source term code package in the ELEBRA MX-850 system

    The implantation of source term code package in the ELEBRA-MX850 system is presented. The source term is formed when radioactive materials generated in nuclear fuel leakage toward containment and the external environment to reactor containment. The implantated version in the ELEBRA system are composed of five codes: MARCH 3, TRAPMELT 3, THCCA, VANESA and NAVA. The original example case was used. The example consists of a small loca accident in a PWR type reactor. A sensitivity study for the TRAPMELT 3 code was carried out, modifying the 'TIME STEP' to estimate the processing time of CPU for executing the original example case. (M.C.K.)

  18. Source term characterization for the radioactive contamination of a river - reservoir system

    The source term for radioactive contamination of the Doamnei-Arges river - reservoir system is more than 90% composed by 58 Co and 60 Co from TRIGA reactors cooling system. Measurable amounts of 137 Cs from Chernobyl accident fallout (1986) is coming into the system from atmosphere (resuspension) and from the catchment area (soil erosion). In this paper, a quantitative analysis of the three contamination pathways is performed. (authors)

  19. Radioactivity source terms for underground engineering application

    The constraints on nuclide production are usually very similar in any underground engineering application of nuclear explosives. However, in some applications the end product could be contaminated unless the proper nuclear device is used. This fact can be illustrated from two underground engineering experiments-Gasbuggy and Sloop. In the Gasbuggy experiment, appreciable tritium has been shown to be present in the gas currently being produced. However, in future gas stimulation applications (as distinct from experiments), a minimum production of tritium by the explosive is desirable since product contamination by this nuclide may place severe limitations on the use of the tritiated gas. In Sloop, where production of copper is the goal of the experiment, product contamination would not be caused by tritium but could result from other nuclides: Thus, gas stimulation could require the use of fission explosives while the lower cost per kiloton of thermonuclear explosives could make them attractive for ore-crushing applications. Because of this consideration, radionuclide production calculations must be made for both fission and for thermonuclear explosives in the underground environment. Such activation calculations materials of construction are performed in a manner similar to that described in another paper, but radionuclide production in the environment must be computed using both fission neutron and 14-MeV neutron sources in order to treat the 'source term' problem realistically. In making such computations, parameter studies including the effects of environmental temperature, neutron shielding, and rock types have been carried out. Results indicate the importance of carefully evaluating the radionuclide production for each individual underground engineering application. (author)

  20. Long-term passive CANDU containment response after a design-basis accident

    A passive CANDU reg-sign containment system, currently being developed, is aimed at limiting the consequences of a postulated accident, by ensuring the structural integrity of the containment building and limiting fission-product release to within siting dose limits, without operator action or reliance on ac power for up to 3 d. All main functions of the containment system, i.e. energy removal, hydrogen mitigation, and fission-product retention, are to be accomplished passively. The passive CANDU containment relies on the passive emergency water system (PEWS) for energy removal after an accident and on passive autocatalytic recombiners (PAR) for hydrogen removal. The key feature of this concept, is a recirculating, buoyancy-driven flow through the recombiners and the tube banks of the PEWS. This paper presents preliminary design calculations for the PEWS tank and tube banks and a simulation of the long-term passive containment response, based on the current CANDU-6 containment, to a large loss-of-coolant/loss-of- emergency coolant injection (LOCA/LOECI) using the GOTHIC code. It is shown that a 1500-M3 PEWS tank, connected to tube banks with a total surface area of 1800 m2, can limit the second pressure peak to about 300 kPa(a) if a recirculating flow is established in the containment building. The PEWS tank water is boiling in the long term, and the peak containment temperature is 114 degrees C. 6 refs., 4 figs

  1. Risk comparisons based on representative source terms with the NUREG-1150 results

    Standardized source terms, based on a specified release of fission products during potential accidents at commercial light water nuclear reactors, have been used for a long time for regulatory purposes. The siting of nuclear power plants, for example, which is governed by Part 100 of the Code of Federal Regulations Title 10, has utilized the source term recommended in TID-14844 supplemented by Regulatory Guides 1.3 and 1.4 and the Standard Review Plan. With the introduction of probabilistic risk assessment (PRA) methods, the source terms became characterized not only by the amount of fission products released, but also by the probability of the release. In the Reactor Safety Study, for example, several categories of source terms, characterized by release severity and probability, were developed for both pressurized and boiling water reactors (PWRs and BWRs). These categories were based on an understanding of the likely paths and associated phenomenology of accident progression following core damage to possible failure of the containment and release to the environment

  2. Workshop on short-term health effects of reactor accidents: Chernobyl

    1986-08-08

    The high-dose early-effects research that has been continued has been done in the context of infrequent accidents with large radiation sources and the use of bone marrow transfusions for treating malignancies, especially leukemia. It thus seemed appropriate to bring together those who have done research on and have had experience with massive whole-body radiation. The objectives were to review what is known about the acute effects of whole-body irradiation, to review the current knowledge of therapy, and particularly of the diagnostic and immunologic problems encountered in bone marrow therapy, and to compare this knowledge with observations made to date on the Chernobyl accident radiation casualties. Dr. Robert Gale, who had helped to care for these casualties, was present at the Workshop. It was hoped that such a review would help those making continuing clinical and pathological observations on the Chernobyl casualties, and that these observations would provide a basis for recommendations for additional research that might result in improved ability to manage successfully this type of severe injury.

  3. Workshop on short-term health effects of reactor accidents: Chernobyl

    The high-dose early-effects research that has been continued has been done in the context of infrequent accidents with large radiation sources and the use of bone marrow transfusions for treating malignancies, especially leukemia. It thus seemed appropriate to bring together those who have done research on and have had experience with massive whole-body radiation. The objectives were to review what is known about the acute effects of whole-body irradiation, to review the current knowledge of therapy, and particularly of the diagnostic and immunologic problems encountered in bone marrow therapy, and to compare this knowledge with observations made to date on the Chernobyl accident radiation casualties. Dr. Robert Gale, who had helped to care for these casualties, was present at the Workshop. It was hoped that such a review would help those making continuing clinical and pathological observations on the Chernobyl casualties, and that these observations would provide a basis for recommendations for additional research that might result in improved ability to manage successfully this type of severe injury

  4. Decision making framework for application of forest countermeasures in the long term after the Chernobyl accident

    After the ChNPP accident a very large part of the territories covered by natural and artificial forests are contaminated with long-lived radionuclides, especially 137Cs. To protect people against exposure associated with forest contamination in the most affected regions of the NIS countries, countermeasures have been developed and recommended for the forest management. The paper presents a decision making framework to optimise forest countermeasures in the long term after the ChNPP accident. The approach presented is based on the analysis of the main exposure pathways and application of radiological, socio-economical and ecological criteria for the selection of optimal countermeasures strategies. Because of the diversity of these criteria modern decision support technologies based on multi-attributive analysis were applied. The results of the application of this approach are presented in a selected study area (Novozybkov district, Bryansk region, Russian Federation). The results prove and emphasize the need for a flexible technique to provide the optimised forest countermeasures taking into account radioecological, social and economic features of contaminated forests

  5. Bremsstrahlung source term estimation for high energy electron accelerators

    Thick target bremsstrahlung source term for 450 MeV and 550 MeV electrons are experimentally determined using booster synchrotron of Indus facility at Raja Ramanna Centre for Advanced Technology, Indore, India. The source term is also simulated using EGSnrc Monte Carlo code. Results from experiment and simulation are found to be in very good agreement. Based on the agreement between experimental and simulated data, the source term is determined up to 3000 MeV by simulation. The paper also describes the studies carried out on the variation of source term when a thin target is considered in place of a thick target, used in earlier studies. - Highlights: • Experimental determination of bremsstrahlung source term at 450 and 550 MeV electrons. • Monte Carlo calculations performed for validation of experimental data. • Thick and thin target bremsstrahlung source term is studied. • Brensstrahlung Source term is determined up to 3 GeV electron energies

  6. Diagnosis and prognosis of the source term by the French Safety Institut during an emergency on a PWR

    The French approach for the diagnosis and the prognosis of the source term during an accident on a PWR is presented and the tools which have been developed to implement this approach at the Institute for Nuclear Protection and Safety (IPSN) are described. (author). 2 refs, 3 figs

  7. SOURCE 2.0: a computer program to calculate fission product release from multiple fuel elements for accident scenarios

    SOURCE 2.0 is a computer code being jointly developed within the Canadian nuclear industry. It will model the necessary mechanisms required to calculate the fission product release for a variety of accident scenarios, including large break loss of coolant accidents with or without emergency coolant injection. This paper presents the origin of SOURCE 2.0, describes the code structure, the fission product mechanisms modelled, and the quality assurance procedures that are being followed during the code's life cycle. (author)

  8. ITER safety task NID-5a: ITER tritium environmental source terms - safety analysis basis

    The Canadian Fusion Fuels Technology Project's (CFFTP) is part of the contribution to ITER task NID-5a, Initial Tritium Source Term. This safety analysis basis constitutes the first part of the work for establishing tritium source terms and is intended to solicit comments and obtain agreement. The analysis objective is to provide an early estimate of tritium environmental source terms for the events to be analyzed. Events that would result in the loss of tritium are: a Loss of Coolant Accident (LOCA), a vacuum vessel boundary breach. a torus exhaust line failure, a fuelling machine process boundary failure, a fuel processing system process boundary failure, a water detritiation system process boundary failure and an isotope separation system process boundary failure. 9 figs

  9. Effect of Fuel Structure Materials on Radiation Source Term in Reactor Core Meltdown

    Jeong, Hae Sun; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The fission product (Radiation Source) releases from the reactor core into the containment is obligatorily evaluated to guarantee the safety of Nuclear Power Plant (NPP) under the hypothetical accident involving a core meltdown. The initial core inventory is used as a starting point of all radiological consequences and effects on the subsequent results of accident assessment. Hence, a proper evaluation for the inventory can be regarded as one of the most important part over the entire procedure of accident analysis. The inventory of fission products is typically evaluated on the basis of the uranium material (e.g., UO2 and USi2) loaded in nuclear fuel assembly, except for the structure materials such as the end fittings, grids, and some kinds of springs. However, the structure materials are continually activated by the neutrons generated from the nuclear fission, and some nuclides of them (e.g., {sup 14}C and {sup 60}Co) can significantly influence on accident assessment. During the severe core accident, the structure components can be also melted with the melting points of temperature relatively lower than uranium material. A series of the calculation were performed by using ORIGEN-S module in SCALE 6.1 package code system. The total activity in each part of structure materials was specifically analyzed from these calculations. The fission product inventory is generally evaluated based on the uranium materials of fuel only, even though the structure components of the assembly are continually activated by the neutrons generated from the nuclear fission. In this study, the activation calculation of the fuel structure materials was performed for the initial source term assessment in the accident of reactor core meltdown. As a result, the lower end fitting and the upper plenum greatly contribute to the total activity except for the cladding material. The nuclides of {sup 56}Mn, '5{sup 1}Cr, {sup 55}Fe, {sup 58}Co, {sup 54}Mn, and {sup 60}Co are analyzed to mainly

  10. Long-term source monitoring with BATSE

    Wilson, R. B.; Harmon, B. A.; Finger, M. H.; Fishman, G. J.; Meegan, C. A.; Paciesas, W. S.

    1992-01-01

    The uncollimated Burst and Transient Source Experiment (BATSE) large area detectors (LADs) are well suited to nearly continuous monitoring of the stronger hard x-ray sources, and time series analysis for pulsars. An overview of the analysis techniques presently being applied to the data are discussed, including representative observations of the Crab Nebula, Crab pulsar, and summaries of the sources detected to data. Results of a search for variability in the Crab Pulsar pulse profile are presented.