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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Evaluation of the Radiological Consequences of Beyond Design Basis Accident for IRT MEPhI

The results of an analysis of the radiological consequences of a beyond design basis accident for IRT MEPhI with high-enrichment uranium fuel and conversion to IRT-3M fuel assemblies with fuel kernel based the alloy U–9% Mo in an aluminum matrix with enrichment 19.7% are presented. The beyond design basis accident is assumed to be blocking of the flow section of the coolant in FA by a foreign object. It was postulated that a single eight-tube FA will melt in this accident. The analysis is performed using approximate conservative estimates. It is shown that the effective dose for the population does not exceed the admissible level. The radiological consequences of the accident for a reactor with high- and low-enrichment fuel are almost indistinguishable.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2010

Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings

ABSTRACT Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 and US NRC Standard Review Plan, Section 3.8) ; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 and 10 CFR 50); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 as well as SECY 90-016, SECY 93-087, and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Reactor Safety for Beyond Design Basis Events

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant's current design basis. The DOE has played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R and D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating as well as better characterize the safety performance of future U.S. nuclear power plants. In pursuing this area of R and D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary end user of the results from this DOE-sponsored work. The purpose of the Reactor Safety Technology R and D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet. The RST R and D activities have evolved from an initial coordinated international effort to assist in the analysis of the Fukushima accident progression and accident response into the following three areas of current work. Fukushima Forensics and Examination Plans: This R and D is focused on providing insights into the actual severe accident progression at Fukushima through planning and interpretation of visual examinations and data collection of in-situ conditions of the damaged units as well as collection and analysis of samples from the damaged reactors. Severe Accident Analyses: This R and D is focused on analyses using existing computer models and their ability to provide information and insights into severe accident progression that aid in the development of severe accident management guidelines (SAMG) and/or training operators on these SAMGs; an auxiliary benefit can be informing improvements in these models. Accident Tolerant Components: This R and D work is focused on analysis or experimental efforts for hardware-related issues, including systems, structures and components with the potential to prevent core degradation or mitigate the effects of beyond-design basis events. This presentation will review current and future R and D activities in this reactor safety area. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Maximal design basis accident of fusion neutron source DEMO-TIN

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Design Basis Document / Owner’s Technical Specification for Nitrate Salt Systems in CSP Projects: Volume 1 Specifications for Parabolic Trough Projects, Volume 2 - Specifications for Central Receiver Projects, Volume 3 - Narrative

Design Basis Documents / Owner’s Technical Specifications are developed for parabolic trough and central receiver power plants using nitrate salt as the heat transport fluid and the thermal storage medium. The goals were to 1) distill the successful experience with nitrate salt systems from as many commercial projects as possible, 2) provide technical bases for equipment design/selection that an owner can impose on an EPC contractor, 3) compile this information in one location to provide guidance on salt systems that is as broadly applicable to as many projects as possible, and 4) work toward an industry consensus on a Design Basis Document that reflects the lessons learned from earlier commercial projects. The document allows an owner to provide design guidance, and to impose a minimum set of requirements, above and beyond those in the normal Codes and Standards, on an EPC contractor in an effort to avoid a repeat of past mistakes. If successful, this would allow salt systems to achieve the same levels of reliability and availability as commercial parabolic trough plants using diphenyl oxide / biphenyl (i.e., Dowtherm A, Therminol VP-1) as the heat transfer fluid. The report consists of 3 volumes: • Volume 1 - Specifications for Parabolic Trough Projects • Volume 2 - Specifications for Central Receiver Projects • Volume 3 - Narrative Volumes 1 and 2 include discussions of the following topics: • Plant functional requirements for the salt systems • Plant operating states, and transitions between states, for the salt systems • Risk analysis of the principal salt components • Plant requirements for the salt systems to meet the functional and the operating requirements • Type of specification for the principal salt components: functional; or prescriptive • Current state of the art for salt systems. Volume 3 discusses a number of cases in which the salt equipment at commercial parabolic trough and central receiver projects has not met the projected levels of reliability and availability. Possible reasons for the sources of the problems are discussed, as is a range of possible alternate designs that could avoid the known problems.

14 SOLAR ENERGY↗
DOE OSTI2016

Transient performance of accident tolerant fuel in design basis accidents

A set of transient analysis has been performed to assess the performance of candidate accident tolerant fuel (ATF) against reference UO{sub 2} fuel with Zircaloy cladding under design basis accident (DBA) conditions. This study examined two ATF options, U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding. The accident analyses were conducted using the system code TRACE for a three-loop PWR plant. PARCS standalone full-core calculations were done for each fuel/cladding combination to provide inputs to the TRACE point-kinetics model. The design basis accidents analyzed were a reactivity insertion accident (RIA) and a large break loss-of-coolant accident (LBLOCA). The RIA was simulated by the insertion of a reactivity ramp of $1 (one dollar) in 0.1 s at hot full power (HFP) condition. The LBLOCA was a simulation of a double-ended pipe break in one of the cold legs of the primary loops. Differences in thermal responses of the fuel and the cladding are attributed to their thermal properties. The U{sub 3}Si{sub 2} fuel with its lower volumetric heat capacity and higher thermal conductivity (i.e. higher thermal diffusivity) generally results in a lower fuel temperature than the UO{sub 2} fuel and a correspondingly faster thermal response. Additional calculations for the U{sub 3}Si{sub 2} fuel shows that in a RIA the cladding material, SiC versus Zr, has little impact on the maximum fuel centerline temperature and the PCT (peak cladding temperature). An increase in the hot rod power factor is shown to raise the centerline fuel temperature in a RIA but has little impact on the PCT. The analysis results show that for a relatively fast decay heat driven transient, such as the LBLOCA, the thermal-hydraulics and fuel/cladding thermal properties dictate the overall response more so than the reactivity feedback. The results show that the PCTs in a LBLOCA are similar for all fuel/cladding combinations considered in this study. The LBLOCA analysis also notes the impact of the asymmetry in the system response. The fuel rods in the sector of the core connected to the hot leg of the broken loop (Sector 1 in the PWR model) is shown to have a much lower PCT than fuel rods in the other five sectors of the core. Results of this study indicate that for the two DBAs considered (RIA and LBLOCA) the transient behavior and the resulting peak cladding temperature of the two ATF options (U{sub 3}Si{sub 2} fuel with SiC cladding and UO{sub 2} fuel with APMT cladding) are similar to the standard light-water reactor fuel (UO{sub 2} fuel with Zircaloy cladding). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

A Deterministic and Probabilistic Framework for Forecasting of Time-Series Damage States and Associated End-of-lIfe of a Pressurizer Water Reactor Surge Line under Design-Basis and Grid-Load-Following Loading Conditions

Argonne National Laboratory in support of the Department of Energy’s Light Water Reactor Sustainability (LWRS) program. The present semi-annual report covers the period between May 2018 and September 2018. In this report, we present the work performed to further improve the capability for structural integrity prediction. Part of this work focused on validating a computational fluid dynamics model for modeling thermal stratification. We also present finite element (FE) model based thermal-mechanical stress analysis of a pressurized water reactor (PWR) surge line under grid-load-following loading condition. In addition, we present work related to end-of-life prediction of a PWR surge line based on ASME code and NUREG-6909 deterministic approaches under design-basis and grid-load-following conditions. Our research in this report period also majorly involved with the development of a strategy for probabilistic fatigue life estimation. First, we applied a Weibull probabilistic modeling approach based on end-of-life fatigue data (traditional S~N data under in-air and PWR water conditions) to estimate the probabilistic life for a given strain amplitude. Second, we developed a data analytics based Markov-Chain-Monte-Carlo (MCMC) model for probabilistic modeling of time-series fatigue damage states and the associated probabilistic end-of-life of a component. Unlike the Weibull type approach, the MCMC type may not require a large number of fatigue tests for estimating the end-of-life for given loading and environmental conditions. We demonstrated the use of this framework through symmetric fatigue tests (stress- or strain-ontrolled tests either under in-air or PWR-water conditions) obtained from our earlier experiments. The MCMC model can use base time-series data either obtained from a model or fatigue experiment. Since experimental data are typically more accurate than model-calculated data, we used experiment-based time-series data to demonstrate the MCMC framework. However, when performing a complex experiment is not possible (due to time and costs involved), a model-based approach can be used to generate the time-series data. Similar model-based data are presented in our earlier report (e.g. for cyclic hardening damage states of a PWR surge line under constant amplitude and design-basis loading [1]). However, for the MCMC work in this report, we used experiment-based time-series damage states obtained from fatigue experiments under design-basis and grid-load-following loading conditions. The experimental loading inputs were based on thermal-mechanical FE model results of a typical PWR surge line. Then, we used the experimental time-series damage state data for forecasting the probabilistic time-series damage states and end-of-life of the PWR surge line. Based on this MCMC modeling strategy, and with the assumed loading, environment and thermal-mechanical boundary conditions, we have estimated that a 316SS-pure base metal PWR surge line would survive at least 159 years with zero failure probability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

The Functional Requirements and Design Basis for Information Barriers

This report summarizes the results of the Information Barrier Working Group workshop held at Sandia National Laboratory in Albuquerque, NM, February 2-4, 1999. This workshop was convened to establish the functional requirements associated with warhead radiation signature information barriers, to identify the major design elements of any such system or approach, and to identify a design basis for each of these major elements. Such information forms the general design basis to be used in designing, fabricating, and evaluating the complete integrated systems developed for specific purposes.

Fuller, James L.↗
DOE OSTI2025

Design Basis Document / Owner’s Technical Specification for Nitrate Salt Systems in CSP Projects (Final Technical Report)

The number of commercial coal, gas, and nuclear projects built over the past 100 years number in the thousands. As such, there is a large database, available to a wide range of commercial engineering contractors, on proven designs. In essence, unsuccessful designs have, through generations of iterations, been identified and then deleted from further consideration. In contrast, the number of commercial parabolic trough projects using nitrate salt for the thermal storage media is perhaps 60. Further, the number of commercial central receiver projects using nitrate salt as the working fluid is on the order of 20, including estimates for China. Given the relative immaturity of salt technology, and commercial pressures to successfully bid new solar projects into a mature electricity market, solar projects often promise more than has been delivered. The Design Basis Document / Owner’s Technical Specification is a first step in the iteration process. The report describes the successful features of commercial projects, outlines a range equipment and system failures in projects that didn’t operate as intended, and provides a draft set of design changes intended to correct the known problems. The product of the study is 3 volumes of technical material; one volume is on parabolic trough technologies; a second is on central receiver technologies, and the third is on potential design changes to parabolic trough and central receiver projects. The 3 volumes, which total some 590 pages, can be found at https://www.solardynllc.com/csp-plant-technologies. One of the principal topics in the report is the use of functional or prescriptive specifications. Functional specifications describe what the equipment needs to do, consistent with the minimum legal requirements of the local jurisdictions. The details of how this is to be accomplished is developed by the engineering contractor. Prescriptive specifications, which are developed by the Owner, prescribe to the engineering contractor how the functional requirements are to be met. This arrangement ensures that the favorable experience from a previous project is repeated. One example is the design code for the hot salt tank in central receiver projects. The closest design basis is API Standard 650 Welded Steel Tanks for Oil Storage. However, the maximum design temperature in API 650 is 260 °C. As such, solar projects have typically adopted a hybrid Code approach, in which allowable material stresses are taken from ASME Section II Materials. Further, since the tanks experience daily changes in temperature and in (static) pressure, and since portions of the tank can operate at stresses beyond the elastic range, the low cycle fatigue life of the tank is conducted using the rules of Section VIII Division 2. However, in a recent study by NREL, the principal damage mechanism was identified as creep rather than fatigue. Further, design stresses permitted under Section VIII Division 2, corresponding to a fatigue life of 30 years, result in projected creep lifetimes of only 2 to 5 years. An alternate design approach, prescribed by the Owner, would be based on Code sections intended for high temperature service in the creep regime. A candidate is Section III Division 5. Granted, this is a nuclear code section, and it’s use would not likely be mandated by local jurisdictions. However, the effects of creep have been deemed to be of sufficient importance that one nuclear project developer, and one central receiver project developer, have stipulated in the tank design specification that the equipment be designed to the requirements of Section III Division 5.

14 SOLAR ENERGY↗
NASA NTRS2009

A Design Basis for Spacecraft Cabin Trace Contaminant Control

Successful trace chemical contamination control is one of the components necessary for achieving good cabin atmospheric quality. While employing seemingly simple process technologies, sizing the active contamination control equipment must employ a reliable design basis for the trace chemical load in the cabin atmosphere. A simplified design basis that draws on experience gained from the International Space Station program is presented. The trace chemical contamination control design load refines generation source magnitudes and includes key chemical functional groups representing both engineering and toxicology challenges.

Perry, Jay L.↗
DOE OSTI2013

Beyond Design Basis Event Pilot Evaluations: Results and Recommendations for Improvements to Enhance Nuclear Safety at DOE Nuclear Facilities (DOE, 2013. A Report to the Secretary of Energy)

In the six months after the March 2011 Fukushima Daiichi nuclear power plant accident in Japan, the U.S. Department of Energy (DOE) took several actions to review the safety of its nuclear facilities and identify situations where near-term improvements could be made. These actions and recommendations were addressed in an August 2011 report to the Secretary of Energy, Review of Requirements and Capabilities for Analyzing and Responding to Beyond Design Basis Events. Based on recommendations in the August 2011 report, DOE embarked on a project to develop and refine guidance that supports improvements in DOE’s processes for analyzing and mitigating beyond design basis events (BDBEs), i.e., events such as earthquakes that are more severe than the events that formed the basis of the design for DOE’s nuclear facilities. The results of this BDBE project and recommendations for further DOE actions are provided in this follow-on report. The main activity of the BDBE project was the pilot application of guidance for evaluation of BDBE analysis and mitigation features at four DOE nuclear facilities representing a range of DOE sites, nuclear facility types/activities, and responsible program offices. The pilot evaluations focused on: (1) BDBE evaluations as documented in the facility Documented Safety Analysis (DSA); (2) potential BDBE vulnerabilities and margins to failure for facility safety features as obtained from general area and specific system walkdowns and design document reviews; and, (3) preparations made in facility and site emergency management programs to respond to severe accidents. The BDBE project also evaluated whether draft BDBE guidance on safety analysis and emergency management could be used to improve the analysis of, and preparations for, mitigating severe accidents and BDBEs. The pilot evaluation team paid close attention to related actions being pursued by the U.S. Nuclear Regulatory Commission, the U.S. commercial nuclear industry, and the international nuclear safety community to help refine the guidance developed for the pilot evaluations.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Limitations of traditional tools for beyond design basis external hazard PRA

Probabilistic risk assessment (PRA) is being used increasingly by the nuclear industry for safety during normal operations as well as for the protection against external hazards. Computation of total risk in an external hazard PRA is dependent on hazard assessment, fragility assessment, and systems analysis. A systems analysis for propagation of component fragilities is conducted using event and fault trees. The event and fault trees for an actual power plant can be fairly large in size, which imposes computational challenges. Hence, certain assumptions are employed for computational efficiency. These assumptions typically represent the conditions imposed during the design basis (DB) scenario. The traditional PRA tools based on these assumptions are also widely applied to perform risk assessment in the context of beyond design basis (BDB) scenarios. However, some of these assumptions may not be valid for certain BDB scenarios. In addition, the probability of dependent failures also increases in BDB scenarios due to common cause failures (CCF) which usually results from design modifications, human errors, etc. In this manuscript, a simple and a relatively more complex illustrative examples are used to show the limitation of these assumptions in the numerical quantification of risk for the case of BDB conditions. Case studies with CCF events across multiple fault trees are also presented to illustrate the effect of these assumptions when traditional approach is used in BDB risk assessment. It is shown that the assumptions are valid for the case of DB conditions but may lead to excessively conservative risk estimates in the case of BDB conditions. Finally, a Bayesian network based top-down algorithm is proposed as an alternative tool for accurate numerical quantification of total risk in systems analysis.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Transportation Package Seal Performance in Beyond Design Basis Thermal Exposures - 12472

The Nuclear Regulatory Commission (NRC) technical report, NUREG/CR-6886, 'Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario', describes, in detail, an evaluation of the potential for a theoretical release of radioactive material from three different spent nuclear fuel (SNF) transportation packages, had they been exposed to the Baltimore tunnel fire that occurred in July of 2001. This evaluation determined the temperatures of various components of the packages, including the seals, using temperatures resulting from models of the Baltimore tunnel fire (as boundary conditions) and finite element models of the SNF packages. For two of the packages evaluated, the analyses indicated that the seals used would have exceeded their continuous-use rated service temperatures, meaning the release of radioactive material could not be ruled out with available information; However, for both of the packages evaluated, the analysis determined, by a bounding calculation, that the maximum potential release was well below the regulatory requirements for releases from a SNF package during the hypothetical accident condition (HAC) sequence of events in 10CFR Part 71. The NRC is investigating the performance of seals in SNF transportation packages exposed to fires that could exceed the HAC fire described in 10CFR Part 71, such as the Baltimore Tunnel Fire that occurred in 2001. The performance of package seals is important for determining the potential release of radioactive material from a package during a beyond-design-basis accident. The seals have lower temperature limits than other package components and are the containment barrier between the environment and the cask contents. The NRC Office of Nuclear Regulatory Research contracted the National Institute of Standards and Technology (NIST) to conduct small-scale thermal testing to obtain experimental data of the performance of seals during extreme temperature exposures. The experimental testing consisted of several small-scale pressure vessels fabricated with a modified ASME flange design and tested metallic and elastomeric seals, similar to those that might be used on an actual SNF transportation package. The vessels were heated in an electrical oven to temperatures as high as 800 deg. C (1472 deg. F), exceeding the rated temperatures of the seals in question. This paper will provide a summary of the testing conducted and present test results and conclusions. Fifteen metallic and elastomeric seal performance tests including two shakedowns were performed under beyond-design-basis thermal exposure conditions; twelve tests used metallic seals, two used ethylene-propylene seals, and one used a Teflon seal. Of the five repeat metallic-seal tests (Tests No.2, 3, 4, 8, and 9), leakage (decreasing vessel pressure) was observed in three of the tests (Tests No.3, 4, and 8) during the 9 h 800 deg. C (1472 deg. F) exposure. The times when the leakage occurred (the vessel pressure started to decrease) varied in the three tests performed. The two shakedown tests were conducted using a 30 min and 4 h exposure to 800 deg. C (1472 deg. F), respectively, and the seal appeared to hold vessel pressure. No leakage (unchanged vessel pressure within the pressure measurement uncertainty) was also observed in the two metallic seal tests that used 100 deg. C incremental heating from 427 deg. C (800 deg. F) to 627 deg. C (1160 deg. F) and from 427 deg. C (800 deg. F) to 727 deg. C (1340 deg. F), respectively, with at least 9 h exposure to each temperature increment. Three repeat metallic seal tests were also conducted at the seal maximum operating temperature of 427 deg. C (800 deg. F) for 9 h. The seal maintained vessel pressure within the measurement of the pressure transducer in all three tests. No leakage was observed in one ethylene-propylene seal tested at 300 deg. C (572 deg. F) for more than 20 hours; however, leakage was found in another ethylene-propylene seal tested at 450 deg. C (842 deg. F) for more than 20 hours (Test No.15). Leakage was also observed in the test (Test No.12) that used a TFE seal subject to 300 deg. C (572 deg. F) exposure. In these fifteen tests the temperature and time exposure variables were changed to map the performance of the metallic and polymeric seals at elevated temperatures beyond the rated operating temperatures. Both the metallic and polymeric seals were able to hold pressure even after leaks were detected and no catastrophic seal failure (i.e., a rapid release of pressure to atmospheric) occurred. Further testing on the polymeric seals would need to be performed to confirm the results in these tests.(authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

LAW Melter Offgas System Design Basis Flowsheets (11/15/2019)

The overall objective of this calculation is to develop design basis flowsheets for confirmation of equipment and piping design associated with the LAW melter offgas system (Systems LOP and LVP). Three flowsheet cases are developed. Minimum, maximum, and nominal offgas cases are developed assuming two melters in operation. The minimum, maximum, and nominal cases are expected during normal operation of the offgas system. An additional minimum case is developed assuming one melter is operating.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Technical Details on Beyond Design Basis Event Pilot Evaluations

The primary focus of the BDBE pilot project was the review of BDBE analysis and mitigation features at four DOE nuclear facilities representing a range of DOE sites, nuclear facility types/activities, and responsible program offices. The pilots looked at (1) how beyond design basis accidents were evaluated and documented in the facility Documented Safety Analysis, (2) potential BDBE vulnerabilities and margins to failure of facility safety features as obtained from general area and specific system walkdowns and design documents reviews, and (3) preparations made in facility and site emergency management programs to respond to severe accidents. It also evaluated whether draft BDBE guidance on safety analysis and emergency management could be used to improve the analysis of and preparations for mitigating severe and beyond design basis accidents. The details of these activities are organized in this report as described below.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Control of Beyond Design Basis Accident in Rbmk with Total Blackout with Prolonged Core Dewatering

A scenario is examined for the development of a beyond design basis accident with loss-of-power for internal needs and functional failures/delays in the actuation of mobile means for emergency cooling RBMK in application to the No. 3 unit of the Kursk NPP. It is shown that actions taken by operating personnel to control pressure in the circulation loop, primarily with the aid of the valves of fast reduction setup for dumping steam in the turbine condensers are effective. The sequence of these actions makes it possible to prevent collapse of the channel tubing at high pressure, preserve the integrity of the reactor space, and effectively limit the accidental emission of radionuclides into the environment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗