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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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12 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

ABWR start-up test analysis using BWR core simulator with three-dimensional direct response matrix method

The ABWR start-up test analysis has been done with the BWR core simulator using the three--dimensional direct response matrix (3D-DRM) method. The Monte Carlo code VMONT made the sub-response matrices for the 3D-DRM method. Each boundary surface was subdivided by 4 x 4 for transverse segments, by 4 for angular segments and by 4 for axial zones in a node. For the calculation speedup, the 3D-DRM code used the divided sub-response matrices data set. The code used the MPI and OpenMP for the parallelized method. The median value is set as the average critical eigenvalues. The changes from the maximum value to the minimum value are 0.34 %{Delta}k with the spectral history method and 0.40 %{Delta}k without it, and the respective standard deviations were 0.12 % and 0.14 %. Using the spectral history method decreased the variation by 0.06 %{Delta}k. The root mean square differences of the axial power distribution were about 6 % between the analysis results and the plant data. Using the currents which converged in the previous exposure step reduced the number of iterations when the CR pattern changed only slightly. The averaged calculation time for each exposure step was about 5 hours on 12 PC Linux cluster servers with Core 2 Quad 3 GHz. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

TRACE/PARCS modelling of rips trip transients for Lungmen ABWR

The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development of advanced BWR fuel bundle with spectral shift rod - BWR core characteristics with SSR

The neutron energy spectrum can be varied during an operation cycle to generate and utilize more plutonium from the non-fissile {sup 238}U by changing the void fraction in the core through control of the core coolant flow rate. This operation method, which is called a spectral shift operation, is practiced in BWRs to save natural uranium. A new component called a spectral shift rod (SSR), which is utilized instead of a conventional water rod, has been introduced to amplify the void fraction change and increase the spectral shift effect. In this study, fuel bundle design with the SSR and core design were carried out for the ABWR and the next generation BWR, HP-ABWR (High-Performance ABWR).The core characteristics with the SSR were evaluated and compared with those when using the conventional water rod. Influences of uncertainty of the water level in the SSR on the safety limit minimum critical power ratio (SLMCPR) were considered for evaluation of the uranium saving effect attained by the SSR. As a result, it was found that the amount of natural uranium needed for an operation cycle could be reduced more than 3% with 20% core coolant flow change and more than 5% with 30% core coolant flow change, in the form of increased discharge exposure by using the SSR compared with the conventional water rod use. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Defining the “proven technology” technical criterion in the reactor technology assessment for Malaysia’s nuclear power program

Developing countries that are considering the deployment of nuclear power plants (NPPs) in the near future need to perform reactor technology assessment (RTA) in order to select the most suitable reactor design. The International Atomic Energy Agency (IAEA) reported in the Common User Considerations (CUC) document that “proven technology” is one of the most important technical criteria for newcomer countries in performing the RTA. The qualitative description of five desired features for “proven technology” is relatively broad and only provides a general guideline to its characterization. This paper proposes a methodology to define the “proven technology” term according to a specific country’s requirements using a three-stage evaluation process. The first evaluation stage screens the available technologies in the market against a predefined minimum Technology Readiness Level (TRL) derived as a condition based on national needs and policy objectives. The result is a list of technology options, which are then assessed in the second evaluation stage against quantitative definitions of CUC desired features for proven technology. The potential technology candidates produced from this evaluation is further narrowed down to obtain a list of proven technology candidates by assessing them against selected risk criteria and the established maximum allowable total score using a scoring matrix. The outcome of this methodology is the proven technology candidates selected using an accurate definition of “proven technology” that fulfills the policy objectives, national needs and risk, and country-specific CUC desired features of the country that performs this assessment. A simplified assessment for Malaysia is carried out to demonstrate and suggest the use of the proposed methodology. In this exercise, ABWR, AP1000, APR1400 and EPR designs assumed the top-ranks of proven technology candidates according to Malaysia’s definition of “proven technology”.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

A step-by-step development of radioactive waste management strategy

Fennovoima is Finland's third nuclear power company founded in 2007. Fennovoima will build a new nuclear power plant, Hanhikivi 1, in Pyhajoki municipality located in Northern Finland. Currently Fennovoima is evaluating bids from reactor suppliers. Originally two suppliers, AREVA and Toshiba Heavy Industries were invited to bid for the plant but later also a bid from Rosatom were invited. The plant supplier will be selected in 2013. Platom is a Finnish company with extensive experience in radioactive waste management. In the beginning of 2008, Platom was assigned by Fennovoima as radioactive waste management consultant and to develop radioactive waste management strategy and to support Fennovoima in negotiations with plant suppliers in waste management aspects. Since Fennovoima has started, there have been some changes to the regulatory requirements which had to be taken into account while developing the strategy. One important change is due to the Government Degree 27.11.2008/736 which establishes a new waste category of very low-level waste following international development and the 'new' IAEA waste classification. Other important change was introduced by the change 342/2008 to the Nuclear Energy Act allowing some nuclear waste to be delivered to another country for treatment. These and many more requirements had to be considered when developing the strategy. Early work by Platom was mostly done to support development of application for Decision-in-Principle. This was the first important licensing step for Fennovoima. Work started with studies and plans which were used by Fennovoima to gather the know-how required to draw up the application. Descriptions of waste streams and waste management technologies were developed as well as preliminary waste inventories and studies for final disposal, including preliminary dimensioning of the repository facilities. Based on these plans nuclear regulator's preliminary safety assessment was performed. The work was successfully completed as in 2010 the Government of Finland granted Fennovoima a permit to build a new nuclear power plant. The work continued in 2012 when bids were received for EPR and ABWR type reactors. Plans developed earlier were elaborated into a detailed strategy and the proposed waste management solutions were evaluated. This allowed feedback from the bids into the strategy and to the bid specification and they both were developed. Also waste inventories were elaborated which allowed development of reactor design specific disposal strategies. The main objective of work has been to establish basis for safe and efficient radioactive waste management which meets all the relevant national and international recommendations, requirements and regulations, takes advantage of best available and state-of-art technologies and offers fit-for-purpose solutions. All the work was performed to accommodate requirement based management system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

100% MOX BWR experimental program design using multi-parameter representative

A new multiparameter representative approach for the design of Advanced full MOX BWR core physics experimental programs is developed. The approach is based on sensitivity analysis of integral parameters to nuclear data, and correlations among different integral parameters. The representativeness method is here used to extract a quantitative relationship between a particular integral response of an experimental mock-up and the same response in a reference project to be designed. The study is applied to the design of the 100% MOX BASALA ABWR experimental program in the EOLE facility. The adopted scheme proposes an original approach to the problem, going from the initial 'microscopic' pin-cells integral parameters to the whole 'macroscopic' assembly integral parameters. This approach enables to collect complementary information necessary to optimize the initial design and to meet target accuracy on the integral parameters to be measured. The study has demonstrated the necessity of new fuel pins fabrication, fulfilling minimal costs requirements, to meet acceptable representativeness on local power distribution. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A defense in depth approach for nuclear power plant accident management

An initiating event may lead to a severe accident if the plant safety functions have been challenged or operators do not follow the appropriate accident management procedures. Beyond design basis accidents are those corresponding to events of very low occurrence probability but such an accident may lead to significant consequences. The defense in depth approach is important to assure nuclear safety even in a severe accident. Plant Damage States (PDS) can be defined by the combination of the possible values for each of the PDS parameters which are showed on the nuclear power plant simulator. PDS is used to identify what the initiating event is, and can also give the information of safety system's status whether they are bypassed, inoperable or not. Initiating event and safety system's status are used in the construction of Containment Event Tree (CET) to determine containment failure modes by using probabilistic risk assessment (PRA) technique. Different initiating events will correspond to different CETs. With these CETs, the core melt frequency of an initiating event can be found. The use of Plant Damage States (PDS) is a symptom-oriented approach. On the other hand, the use of Containment Event Tree (CET) is an event-oriented approach. In this study, the Taiwan's fourth nuclear power plants, the Lungmen nuclear power station (LNPS), which is an advanced boiling water reactor (ABWR) with fully digitized instrumentation and control (I and C) system is chosen as the target plant. The LNPS full scope engineering simulator is used to generate the testing data for method development. The following common initiating events are considered in this study: loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), loss of offsite power (LOOP), station blackout (SBO). Studies have indicated that the combination of the symptom-oriented approach and the event-oriented approach can be helpful to find mitigation strategies and is useful for the accident management. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Safety demonstration of a class 1 smart device - 283

Horizon Nuclear Power intends to build Advanced Boiling Water Reactors (ABWR) at Wylfa and Oldbury in the UK, based on the Hitachi design. In accordance with UK policy for new nuclear build, Hitachi, as the reactor designer, is the requesting party to the Generic Design Assessment (GDA) during which the reactor design will be reviewed by the Office for Nuclear Regulation (ONR) and the Environment Agency. An important step in the GDA process is to demonstrate the viability of the approach developed by Hitachi-GE for the assessment and justification of smart devices. This was done by means of pilot studies of Safety Class (SC) 1 and SC2 devices. This paper will describe the scope, criteria, process and approach for the SC1 pilot study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Improved Core Design of the Supercritical Water-cooled Reactor CSR1000

The SCWR (Supercritical water-cooled reactor) is one of the Generation IV reactor concepts. The economic performance of the SCWR has a core focus, and in this study forms one of the key design goals. Based on the technology of ABWR and PWR, a conceptual design of CSR1000 (China Supercritical water-cooled Reactor with the rated electric power of 1000 MWe) core was proposed with features such as single water rod, combined square fuel assembly (FA),two flow-pass of coolant in core, high-leakage reload core with 157 FAs, burn-up of 350 EFPD. The refueling cycle of the CSR1000 design was about 12-months, which is much shorter than the generation 3 PWR of 18-months. Thus, it is necessary to enhance the economic performance of CSR1000 by prolonging the refueling cycle. Based on conceptual design of the CSR1000, the FA and the core design are improved for prolonging the refueling cycle of CSR1000. This paper represents an important reference and technical reserve for further SCWR core design study.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗