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Sample records for water breeder reactor

  1. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  2. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  3. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  4. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  5. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  6. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  7. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  8. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    Energy Technology Data Exchange (ETDEWEB)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  9. Breeder Reactors, Understanding the Atom Series.

    Science.gov (United States)

    Mitchell, Walter, III; Turner, Stanley E.

    The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

  10. Fast breeder reactor protection system

    Science.gov (United States)

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  11. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  12. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  13. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  14. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  15. Multiple recycling of fuel in prototype fast breeder reactor in a closed fuel cycle with pressurized heavy-water reactor external feed

    Indian Academy of Sciences (India)

    G Pandikumar; A John Arul; P Puthiyavinayagam; P Chellapandi

    2015-10-01

    A fast breeder reactor (FBR) closed fuel cycle involves recycling of the discharged fuel, after reprocessing and refabrication, in order to utilize the unburnt fuel and the bred fissile material. Our previous study in this regard for the prototype fast breeder reactor (PFBR) indicated the possibility of multiple recycling with self-sufficiency. It was found that the change in Pu composition becomes negligible (less than 1%) after a few cycles. The core-1 Pu increases by 3% from the beginning of cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th by only 0.3%. In this work, the possibility of multiple recycling of PFBR fuel with external plutonium feed from pressurized heavy-water reactor (PHWR) is examined. Modified in-core cooling and reprocessing periods are considered. The impact of multiple recycling on PFBR core physics parameters due to the changes in the fuel composition has been brought out. Instead of separate recovery considered for the core and axial blankets in the earlier studies, combined fuel recovery is considered in this study. With these modifications and also with PHWR Pu as external feed, the study on PFBR fuel recycling is repeated. It is observed that the core-1 initial Pu inventory increases by 3.5% from cycle-0 to that of recycle-1, the Pu increase from the beginning of the 9th cycle to that of the 10th is only 0.35%. A comparison of the studies done with different external plutonium options viz., PHWR and PFBR radial blanket has also been made.

  16. Dynamic model of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaidyanathan, G., E-mail: vaidya@igcar.gov.i [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2010-04-15

    Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.

  17. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  18. Safeguards in prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Deshimaru, Takehide; Tomura, Katsuji; Okuda, Yosihisa; Iwamoto, Tomonori [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1994-12-31

    MONJU is the prototype fast breeder reactor in Japan designed to have the electricity output of 280 MWe. Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga site. The loading of the core fuel assemblies to the core have been started since October 1993 and the pre-operational test is undergoing. MONJU uses 198 MOX fuel assemblies as core fuel and 172 DU assemblies as blanket fuel. Assemblies loaded in core and stored in the ex-vessel storage tank (EVST) exist in liquid sodium. These Pu containing fuel assemblies, MOX and irradiated DU, are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area are requested to be verified. The verification of the flows is designed to be made with fuel flow monitors measuring radiations, which can abridge the inspector attendance during the fuel handling. This paper describes the detailed aspects of the fuel transfers in MONJU facility and the verification of them through flow monitors together with the functions of other safeguards equipments. (author).

  19. Special Analysis for the Disposal of the INL Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) Waste Stream at the Area 5 Radioactive Waste Management Site

    Energy Technology Data Exchange (ETDEWEB)

    Shott, Gregory [National Security Technologies, LLC, Las Vegas, NV (United States)

    2017-03-21

    This special analysis (SA) evaluates whether the Idaho National Laboratory (INL) Waste Associated with the Unirradiated Light Water Breeder Reactor (LWBR) waste stream (INEL167203QR1, Revision 0) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INL Waste Associated with the Unirradiated LWBR waste meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P performance objectives (DOE 1999). The INL Waste Associated with the Unirradiated LWBR waste stream is recommended for acceptance with the condition that the total uranium-233 (233U) inventory be limited to 2.7E13 Bq (7.2E2 Ci).

  20. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  1. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    P K Vijayan; A Basak; I V Dulera; K K Vaze; S Basu; R K Sinha

    2015-09-01

    The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th-based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

  2. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

    2016-02-15

    Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

  3. Safety aspects of fuel behaviour during faults and accidents in pressurised water reactors and in liquid sodium cooled fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H. (UKAEA Information Services Branch, London); Matthews, J.R. (UKAEA Harwell Lab. (UK). Theoretical Physics Div.); Potter, P.E. (UKAEA Harwell Lab. (UK). Chemistry Div.)

    1989-07-01

    The good safety record of electrical power generating reactors in the European Community is based on a substantial effort to understand the safety characteristics of the reactors and their fuel. In this paper the present state of knowledge of oxide fuels used in current European reactors is reviewed. The main theme of the paper is the importance of the role of fission products and the chemical state of the fuel on all aspects of fuel behaviour. The paper is split into two parts. The first part deals with those aspects specific to water reactors using UO{sub 2} based fuels. The second part of the paper deals with mixed-oxide fuels and the sodium cooled reactors. In each part the following aspects are described: Chemical constitution of the fuel; fuel performance and failure limits; failed fuel behaviour; fuel behaviour in accidents; and the interactions in degraded cores after hypothetical accidents. Future directions of safety related fuel work in Europe are identified. (orig.).

  4. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  5. Fission-suppressed hybrid reactor: the fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  6. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    Science.gov (United States)

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  7. ORIGEN2 model and results for the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A G; Bjerke, M A

    1982-06-01

    Reactor physics calculations and literature information acquisition have led to the development of a Clinch River Breeder Reactor (CRBR) model for the ORIGEN2 computer code. The model is based on cross sections taken directly from physics codes. Details are presented concerning the physical description of the fuel assemblies, the fuel management scheme, irradiation parameters, and initial material compositions. The ORIGEN2 model for the CRBR has been implemented, resulting in the production of graphical and tabular characteristics (radioactivity, thermal power, and toxicity) of CRBR spent fuel, high-level waste, and fuel-assembly structural material waste as a function of decay time. Characteristics for pressurized water reactors (PWRs), commercial liquid-metal fast breeder reactors (LMFBRs), and the Fast Flux Test Facility (FFTF) have also been included in this report for comparison with the CRBR data.

  8. Multiple recycling of fuel in prototype fast breeder reactor

    Indian Academy of Sciences (India)

    G Pandikumar; V Gopalakrishnan; P Mohanakrishnan

    2009-05-01

    In a thermal neutron reactor, multiple recycle of U–Pu fuel is not possible due to degradation of fissile content of Pu in just one recycle. In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder reactor under construction at Kalpakkam. After about five recycles, the cycle-to-cycle variation in the above parameters is below 1%.

  9. Ultrasonic decontamination of prototype fast breeder reactor fuel pins.

    Science.gov (United States)

    Kumar, Aniruddha; Bhatt, R B; Behere, P G; Afzal, Mohd

    2014-04-01

    Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%.

  10. Molten Salt Breeder Reactor Analysis Methods

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinsu; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Utilizing the uranium-thorium fuel cycle shows considerable potential for the possibility of MSR. The concept of MSBR should be revised because of molten salt reactor's advantage such as outstanding neutron economy, possibility of continuous online reprocessing and refueling, a high level of inherent safety, and economic benefit by keeping off the fuel fabrication process. For the development of MSR research, this paper provides the MSBR single-cell, two-cell and whole core model for computer code input, and several calculation results including depletion calculation of each models. The calculations are carried out by using MCNP6, a Monte Carlo computer code, which has CINDER90 for depletion calculation using ENDF-VII nuclear data. From the calculation results of various reactor design parameters, the temperature coefficients are all negative at the initial state and MTC becomes positive at the equilibrium state. From the results of core rod worth, the graphite control rod alone cannot makes the core subcritical at initial state. But the equilibrium state, the core can be made subcritical state only by graphite control rods. Through the comparison of the results of each models, the two-cell method can represent the MSBR core model more accurately with a little more computational resources than the single-cell method. Many of the thermal spectrum MSR have adopted a multi-region single-fluid strategy.

  11. Decommissioning of Experimental Breeder Reactor - II Complex, Post Sodium Draining

    Energy Technology Data Exchange (ETDEWEB)

    J. A. (Bart) Michelbacher; S. Paul Henslee; Collin J. Knight; Steven R. sherman

    2005-09-01

    The Experimental Breeder Reactor - II (EBR-II) was shutdown in September 1994 as mandated by the United States Department of Energy. This sodium-cooled reactor had been in service since 1964. The bulk sodium was drained from the primary and secondary systems and processed. Residual sodium remaining in the systems after draining was converted into sodium bicarbonate using humid carbon dioxide. This technique was tested at Argonne National Laboratory in Illinois under controlled conditions, then demonstrated on a larger scale by treating residual sodium within the EBR-II secondary cooling system, followed by the primary tank. This process, terminated in 2002, was used to place a layer of sodium bicarbonate over all exposed surfaces of sodium. Treatment of the remaining EBR-II sodium is governed by the Resource Conservation and Recovery Act (RCRA). The Idaho Department of Environmental Quality issued a RCRA Operating Permit in 2002, mandating that all hazardous materials be removed from EBR-II within a 10 year period, with the ability to extend the permit and treatment period for another 10 years. A preliminary plan has been formulated to remove the remaining sodium and NaK from the primary and secondary systems using moist carbon dioxide, steam and nitrogen, and a water flush. The moist carbon dioxide treatment was resumed in May 2004. As of August 2005, approximately 60% of the residual sodium within the EBR-II primary tank had been treated. This process will continue through the end of 2005, when it is forecast that the process will become increasingly ineffective. At that time, subsequent treatment processes will be planned and initiated. It should be noted that the processes and anticipated costs associated with these processes are preliminary. Detailed engineering has not been performed, and approval for these methods has not been obtained from the regulator or the sponsors.

  12. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  13. Development of fuels and structural materials for fast breeder reactors

    Indian Academy of Sciences (India)

    Baldev Raj; S L Mannan; P R Vasudeva Rao; M D Mathew

    2002-10-01

    Fast breeder reactors (FBRs) are destined to play a crucial role inthe Indian nuclear power programme in the foreseeable future. FBR technology involves a multi-disciplinary approach to solve the various challenges in the areas of fuel and materials development. Fuels for FBRs have significantly higher concentration of fissile material than in thermal reactors, with a matching increase in burn-up. The design of the fuel is an important aspect which has to be optimised for efficient, economic and safe production of power. FBR components operate under hostile and demanding environment of high neutron flux, liquid sodium coolant and elevated temperatures. Resistance to void swelling, irradiation creep, and irradiation embrittlement are therefore major considerations in the choice of materials for the core components. Structural and steam generator materials should have good resistance to creep, low cycle fatigue, creep-fatigue interaction and sodium corrosion. The development of carbide fuel and structural materials for the Fast Breeder Test Reactor at Kalpakkam was a great technological challenge. At the Indira Gandhi Centre for Atomic Research (IGCAR), advanced research facilities have been established, and extensive studies have been carried out in the areas of fuel and materials development. This has laid the foundation for the design and development of a 500 MWe Prototype Fast Breeder Reactor. Highlights of some of these studies are discussed in this paper in the context of our mission to develop and deploy FBR technology for the energy security of India in the 21st century.

  14. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  15. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  16. Safeguards in the prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Usami, S.; Deshimaru, T.; Tomura, K. [Power Reactor and Nuclear Fuels Development Corporation, Ibaraki-ken (Japan)

    1995-12-31

    MONJU is a prototype fast breeder reactor in Japan designed to have a 280-MW(electric) output. The Power Reactor and Nuclear Fuel Development Corporation (PNC) started its construction in the autumn of 1985 in Tsuruga. The loading of the core fuel assemblies was started in October 1993, and the preoperational test is ongoing. MONJU uses 198 mixed-oxide (MOX) fuel assemblies as core fuel and 172 depleted uranium assemblies as blanket fuel. Assemblies loaded in-core and stored in the ex-vessel storage tank (EVST) reside in liquid sodium. These plutonium-containing fuel assemblies, MOX, and irradiated depleted uranium are regarded as in the difficult-to-access area, and the flows of fuel assemblies into and out of the area must be verified. Flow is verified by fuel flow monitors measuring radiation, which can limit inspector attendance during fuel handling.

  17. Risk-assessment methodology for fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ott, K. O.

    1976-04-01

    The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed.

  18. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  19. Impact of radionuclides on maintenance of Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Olson, W.H.

    1985-01-01

    More than 20 years of Experimental-Breeder-Reactor-II (EBR-II) operation has demonstrated the capability to maintain radioactive equipment without undue radiation exposure to operating and maintenance personnel. The dominant radioisotopes in EBR-II primary systems are the activated corrosion product /sup 54/Mn and the fission products /sup 90/Sr and /sup 137/Cs. The presence of radioisotopes from direct activation, deposit of activated corrosion products, and release of fission products from breached fuel elements dictates special procedures, equipment, and planning but does not prohibit maintenance activities. Since 1977, the average yearly exposure of operating and maintenance personnel has been reduced while the radioactivity of systems and components has increased.

  20. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    Shott, Gregory [NSTec

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  1. A vision of inexhaustible energy: The fast breeder reactor in Swedish nuclear power history 1945-80; Visionen om outtoemlig energi: Bridreaktorn i svensk kaernkraftshistoria 1945-80

    Energy Technology Data Exchange (ETDEWEB)

    Fjaestad, Maja, E-mail: majaf@k_th.se

    2010-03-15

    The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and 1960s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy, and thereby connected it to utopian ideas about an eternal supply of energy, Furthermore. the ideas of breeder reactors were a vital part of the post-war visions about the nuclear future. This dissertation investigates the plans for breeder reactors in Sweden, connecting them to the contemporary development of nuclear power with heavy or light water and the discussions of nuclear weapons, as well as to the general visions of a prosperous technological future. The history of the Swedish breeder reactor is traced from high hopes in the beginning, via the fiasco of the Swedish heavy water program, partly focusing on the activities at the company AB Atomenergi and investigating how it planned and argued for its breeder program and how this was received by the politicians. The story continues into the intensive environmental movement in the 1970s, ending with the Swedish referendum on nuclear energy in 1980, which can be seen as the final point for the Swedish breeder. The thesis discusses how the nuclear breeder reactor was transformed from an argument for nuclear power to an argument against it. The breeder began as a part of the vision of a society with abundant energy, but was later seen as a threat against the new sustainable world. The nuclear breeder reactor is an example of a technological vision that did not meet its industrial expectations. But that does not prevent the fact that breeder was an influential technology in an age where important decisions about nuclear energy were made. The thesis argues that important decisions about the contemporary reactors were taken with the idea that they in a foreseeable future would be replaced with the efficient breeder. And the last word on the breeder reactor is not said - today, reactor engineers around the world are

  2. The Case Against the Fast Breeder Reactor: An Anti-Nuclear Establishment View.

    Science.gov (United States)

    Lovins, Amory B.

    1973-01-01

    Environmentalists lobby points out that hazards which may result from mistakes in proposed fast breeder reactor for additional energy can be detrimental for mankind. Such projects must be carefully planned and cautiously executed. (PS)

  3. Slow and stunted: Plutonium accounting and the growth of fast breeder reactors in India

    Energy Technology Data Exchange (ETDEWEB)

    Ramana, M.V. [Program in Science, Technology and Environmental Policy and Program on Science and Global Security, Woodrow Wilson School of Public and International Affairs, Princeton University, Princeton, NJ 08542 (United States); Suchitra, J.Y. [Independent Researcher, 303, Meenakshi Apartments, 45 Ranga Rao Road, Shankarapuram, Bangalore 560004 (India)

    2009-12-15

    The Indian Department of Atomic Energy (DAE) has projected a large growth of nuclear power in the country predominantly based on breeder reactors. These projections use a simplistic methodology that does not carefully account for the availability of plutonium that is required to fuel breeder reactors. In this paper, we demonstrate that this methodology is problematic, in particular that it would result in negative balances of plutonium if the DAE's projections were to come true. The DAE's projections also ignore constraints coming from reprocessing capacity in the country. As an alternative, we project the possible growth of nuclear power based on breeder reactors using a methodology consistent with plutonium constraints. The resulting breeder reactor capacity will be only between 17% and 40% of the DAE's projections, and will likely never constitute a major source of electricity in India for several decades at the very least. (author)

  4. EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs

    Energy Technology Data Exchange (ETDEWEB)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

  5. Progress in studies of Li/sub 17/Pb/sub 83/ as liquid breeder for fusion reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.

    1983-09-01

    A review of the experimental and conceptual design work in progress at JRC-Ispra to investigate the feasibility of the eutectic Li/sub 17/Pb/sub 83/ as a liquid breeder for experimental power reactors is presented. Results of recent measurements to implement the data base of this material are given in the following areas: physical parameters, hydrogen solubility and recovery, chemical reactivity with air and water, compatibility with steel. The studies carried out on blanket concepts for the INTOR (International Tokamak Reactor)/NET (Next European Torus) projects are outlined and discussed.

  6. Determination and correlation of mass transfer coefficients in a stirred cell. [Molten Salt Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, J.; Bloxom, S.R.; Keeler, J.B.; Roth, S.R.

    1975-12-17

    In the proposed Molten Salt Breeder Reactor flowsheet, a fraction of the rare earth fission products is removed from the fuel salt in mass transfer cells. To obtain design parameters for this extraction, the effect of cell size, blade diameter, phase volume, and agitation rate on the mass transfer for a high density ratio system (mercury/water) in nondispersing square cross section contactors was determined. Aqueous side mass transfer coefficients were measured by polarography over a wide range of operating conditions. Correlations for the experimental mass transfer coefficients as functions of the operating parameters are presented. Several techniques for measuring mercury-side mass transfer coefficients were evaluated and a new one is recommended. (auth)

  7. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  8. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.W.; Baes, C.F. III; Dunning, D.E. Jr.

    1980-05-01

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections.

  9. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  10. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

  11. Effect of sodium environment on breeder reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Eichelberger, R.L.; Brehm, W.F.

    1978-01-01

    This paper reviews the interaction phenomena between sodium and reactor materials, and the application to reactor design and operation. Both corrosion and deposition are considered. Sodium removal and decontamination is briefly mentioned, together with some recent developments that are still in the experimental stage. No new theories of sodium-material interaction behavior are offered. 56 references.

  12. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  13. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  14. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Padmakumar, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)], E-mail: gpk@igcar.gov.in; Pandey, G.K.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2009-06-15

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  15. Liquid-metal pumps for large-scale breeder-reactor plant (prototype pump)

    Energy Technology Data Exchange (ETDEWEB)

    Lindsay, M. (comp.)

    1976-07-01

    This report presents the recommended pump design for use in Large Scale Liquid Metal Fast Breeder Reactor plants. The base design for the pump will circulate 127,000 GPM of liquid sodium at temperatures up to 850/sup 0/F and with a total discharge head at the design point of 500 feet Na with an impeller that is 40 feet below the sodium seal. The pump design is predicated on developing an impeller design which will have a suction specific speed (S/sub n/) of about 20,000 with 20 feet NPSH available, which will result in a pump speed of 530 RPM at design conditions. The design is based on the technology developed in the design and fabrication of FFTF pumps, the design efforts for the Clinch River Breeder Reactor Pump design study and other technology.

  16. Comparison of containment systems for large sodium-cooled breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Seeman, S.E.; Armstrong, G.R.

    1978-04-01

    Three types of containment configurations for large breeder reactors (1000 MWe/2500 MWt) were compared to determine characteristic responses of each type to selected hypothetical accident conditions. The three types of system configurations were the single, double and containment-confinement designs. Two types of postulated hypothetical accidents were applied as a basis for evaluation--the energetic core disassambly accident and the whole core melt-through accident. More than 20 separate cases were analyzed in which design and accident assumptions were varied. Based on this study the containment-confinement configuration is shown to be the most promising for limiting off-site radiation doses for large sodium-cooled breeder reactors given the specific assumptions on the initial accident conditions.

  17. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  18. Assessment of gel-sphere-pac fuel for fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lackey, W J; Selle, J E [comps.

    1978-10-01

    An assessment of the state of the art for the gel-sphere-pac process was undertaken to provide a sound basis for further development of the technology. Information is provided on sol preparation, sphere forming, drying, sintering, characterization, loading, fuel rod inspection, and irradiation performance. In addition, discussions are included on: evaluation of the potential for scale-up to production capacities, potential problems associated with remote operation, and future work required to further develop the technology. Three techniques are available for microsphere production: (1) internal gelation, (2) external gelation, and (3) gelation by water extraction. Each has its own advantages and disadvantages; for example, internal gelation appears better suited to the preparation of large spheres than the other processes. Numerous advantages and disadvantages are discussed in detail. Scale-up or remote operation of these techniques appears achievable, although some would require less development than others. Techniques have been developed for drying and sintering spheres. Extensive technology has been developed for sphere characterization, handling, and the loading and inspection of fuel pins. Data available to date indicates that sphere-pac oxide fuel will perform similarly to pellet oxide fuels under fast breeder reactor operating conditions. Gel-sphere-pac technology also appears attractive for carbide fuels.

  19. Cellular convection in vertical annuli of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hemanath, M.G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)], E-mail: hemanath@igcar.gov.in; Meikandamurthy, C.; Ramakrishnan, V.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2007-08-15

    In the pool type fast reactors the roof structure is penetrated by a number of pumps and heat exchangers that are cylindrical in shape. Sandwiched between the free surface of sodium and the roof structure, is stagnant argon gas, which can flow in the annular space between the components and roof structure, as a thermosyphon. These thermosyphons not only transport heat from sodium to roof structure, but also result in cellular convection in vertical annuli resulting in circumferential temperature asymmetry of the penetrating components. There is need to know the temperature asymmetry as it can cause tilting of the components. Experiments were carried out in an annulus model to predict the circumferential temperature difference with and without sodium in the test vessel. Three-dimensional analysis was also carried out using PHOENICS CFD code and compared with the experiment. This paper describes the experimental details, the theoretical analysis and their comparison.

  20. Choice of rotatable plug seals for prototype fast breeder reactor: Review of historical perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev, E-mail: baldev.dr@gmail.com

    2015-09-15

    Highlights: • Choice and arrangement of elastomeric inflatable and backup seals as primary and secondary barriers. • With survey (mid-1930s onwards) of reactor, sealing, R&D and rubber technology. • Load, reliability, safety, life and economy of seals and reactors are key factors. • PFBR blends concepts and experience of MOX fuelled FBRs with original solutions. • R&D indicates inflatable seal advanced fluoroelastomer pivotal in unifying nuclear sealing. - Abstract: Choice and arrangement of elastomeric primary inflatable and secondary backup seals for the rotatable plugs (RPs) of 500 MW (e), sodium cooled, pool type, 2-loop, mixed oxide (MOX) fuelled Prototype Fast Breeder Reactor (PFBR) is depicted with review of various historical perspectives. Static and dynamic operation, largest diameters (PFBR: ∼6.4 m, ∼4.2 m), widest gaps and variations (5 ± 2 mm) and demanding operating requirements make RP openings on top shield (TS) the most difficult to seal which necessitated extensive development from 1950s to early 1990s. Liquid metal freeze seals with life equivalent to reactor prevailed as primary barrier (France, Japan, U.S.S.R.) during pre-1980s in spite of bulk, cost and complexity due to the abilities to meet zero leakage and resist core disruptive accident (CDA). Redefinition of CDA as beyond design basis accident, tolerable leakage and enhanced economisation drive during post-1980s established elastomeric inflatable seal as primary barrier excepting in U.S.S.R. (MOX fuel, freeze seal) and U.S.A. (metallic fuel). Choice of inflatable seal for PFBR RPs considers these perspectives, inherent advantages of elastomers and those of inflatable seals which maximise seal life. Choice of elastomeric backup seal as secondary barrier was governed by reliability and minimisation as well as distribution of load (temperature, radiation, mist) to maximise seal life. The compact sealing combination brings the hanging RPs at about the same elevation to reduce

  1. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  2. The fast breeder reactor Rapsodie (1962); Le reacteur rapide surregenerateur rapsodie (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Vautrey, L.; Zaleski, C.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1962-07-01

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors) [French] Dans ce rapport, les auteurs font le point du projet RAPSODIE (reacteur francais surregenerateur a neutrons rapides), au moment du debut effectif de sa construction. On y trouvera decrits: les principales caracteristiques neutroniques et thermiques, le bloc pile et les circuits de refroidissement, les principaux moyens de manutention des ensembles actifs ou contamines, les principes et les moyens qui regissent la conduite du reacteur, les fonctions et l'implantation des divers batiments. La divergence de RAPSODIE est prevue pour 1964. (auteurs)

  3. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  4. Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  5. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  6. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This final report is a complete conceptual design study of a mechanical pump for a large scale breeder reactor plant. The pumps are located in the cold leg side of the loops. This makes the net positive suction head available - NPSHA - low, and is, in fact, a major influencing factor in the design. Where possible, experience gained from the Clinch River Project and the FFTF is used in this study. Experience gained in the design, manufacturer, and testing of pumps in general and sodium pumps in particular is reflected in this report. The report includes estimated cost and time schedule for design, manufacture, and testing. It also includes a recommendation for development needs.

  7. Deterioration of limestone aggregate mortars by liquid sodium in fast breeder reactor environment

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed Haneefa, K., E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F.C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2014-08-15

    Highlights: • Limestone mortars were exposed to liquid sodium exposure at 550 °C. • Micro-analytical techniques were used to characterize the exposed specimens. • The performance of limestone mortar was greatly influenced by w/c. • The fundamental degradation mechanisms of limestone mortars were identified. - Abstract: Hot liquid sodium at 550 °C can interact with concrete in the scenario of an accidental spillage of sodium in liquid metal cooled fast breeder reactors. To protect the structural concrete from thermo-chemical degradation, a sacrificial layer of limestone aggregate concrete is provided over it. This study investigates the fundamental mechanisms of thermo-chemical interaction between the hot liquid sodium and limestone mortars at 550 °C for a duration of 30 min in open air. The investigation involves four different types of cement with variation of water-to-cement ratios (w/c) from 0.4 to 0.6. Comprehensive analysis of experimental results reveals that the degree of damage experienced by limestone mortars displayed an upward trend with increase in w/c ratios for a given type of cement. Performance of fly ash based Portland pozzolana cement was superior to other types of cements for a w/c of 0.55. The fundamental degradation mechanisms of limestone mortars during hot liquid sodium interactions include alterations in cement paste phase, formation of sodium compounds from the interaction between solid phases of cement paste and aggregate, modifications of interfacial transition zone (ITZ), decomposition of CaCO{sub 3}, widening and etching of rhombohedral cleavages, and subsequent breaking through the weakest rhombohedral cleavage planes of calcite, staining, ferric oxidation in grain boundaries and disintegration of impurity minerals in limestone.

  8. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  9. Development of fuel flow monitoring system in prototype fast breeder reactor 'MONJU'

    Energy Technology Data Exchange (ETDEWEB)

    Tomura, Katsuji; Deshimaru; Takehide; Okuda, Yoshihisa; Ohba, Toshio (Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office); Ishikawa, Kouichi

    1994-06-01

    A new safeguards approach of Prototype Fast Breeder Reactor 'MONJU' has been studied by Japanese Government, IAEA and PNC to meet 1991-1995 safeguards criteria. As the result, a fuel flow monitoring system has been introduced in 'MONJU'. Development of the system has been conducted by PNC and IAEA with technical support of Los Alamos National Laboratory. Safeguards measures in unattended mode with the system can detect fuel loading and unloading into and from the reactor core and distinguish what kind of the fuel. The system are consisted of three monitors using neutron and gamma-ray measurements and video surveillance system. Installation of these monitors was finished by PNC and acceptance test by Japanese Government and IAEA was carried out March, 1992. (author).

  10. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  11. The Early Characterization of Irradiation Effects in Stainless Steels at the Experimental Breeder Reactor-II

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Porter

    2008-01-01

    The new Global Nuclear Energy Partnership (GNEP) program is revitalizing interest in materials development for fast spectrum reactors. With this comes the need for new, high-performance materials that are resistant to property changes caused by radiation damage. In the 1970s there was an effort to monitor the irradiation effects on stainless steels used in fast reactor cores, largely because there were a number of ‘surprises’ where materials subjected to a high flux of fast neutrons incurred dimensional and property changes that had not been expected. In the U.S., this applied to the Experimental Breeder Reactor-II. Void swelling and irradiation-induced creep caused dimensional changes in the reactor components that shortened their useful lifetime and impacted reactor operations by creating fuel handling difficulties and reactivity anomalies. The surveillance programs and early experiments studied the simplest of austenitic stainless steels, such as Types 304 and 304L stainless steel, and led to some basic understanding of the links between these irradiation effects and microchemical changes within the steel caused by operational variables such as temperature, neutron flux and neutron fluence. Some of the observations helped to define later alloy development programs designed to produce alloys that were much more resistant to the effects of neutron irradiation.

  12. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    Energy Technology Data Exchange (ETDEWEB)

    Harto, Andang Widi [Engineering Physics Department, Faculty of Engineering, Gadjah Mada University (Indonesia)

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  13. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rajan Babu, V., E-mail: vrb@igcar.gov.i [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India); Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603 102 (India)

    2010-07-15

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO{sub 2} process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  14. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    Science.gov (United States)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  15. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    Science.gov (United States)

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  16. Distribution of liquid sodium in the inlet plenum of steam generator in a Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Patil, Laxman T. [Department of Chemical Engineering, Institute of Chemical Technology, N. M. Parikh Marg, Matunga, Mumbai 400019 (India); Patwardhan, A.W., E-mail: awp@udct.or [Department of Chemical Engineering, Institute of Chemical Technology, N. M. Parikh Marg, Matunga, Mumbai 400019 (India); Padmakumar, G.; Vaidyanathan, G. [Experimental Thermal Hydraulics Section, Separation Technology and Hydraulics Division, Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2010-04-15

    Experimental and Computational Fluid Dynamics (CFD) investigations have been carried out on a 1/5th scale model of the inlet plenum of steam generator (SG) used in the Fast Breeder Reactor (FBR) technology. The distribution of liquid sodium in the inlet plenum of the steam generator strongly affects the thermal as well as mechanical performance of the steam generator. In the present work, flow distribution in a scaled down model has been investigated. Various strategies adopted for obtaining uniform flow distribution have been evaluated. Experiments have been conducted to measure the axial and radial velocity distributions using Ultrasonic Velocity Profiler (UVP) under a variety of geometries. Computational Fluid Dynamics (CFD) studies have been carried out for various geometries. On the basis of these experiments and CFD simulations, various flow distribution devices have been compared.

  17. Compendium of computer codes for the safety analysis of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    1977-10-01

    The objective of the compendium is to provide the reader with a guide which briefly describes many of the computer codes used for liquid metal fast breeder reactor safety analyses, since it is for this system that most of the codes have been developed. The compendium is designed to address the following frequently asked questions from individuals in licensing and research and development activities: (1) What does the code do. (2) To what safety problems has it been applied. (3) What are the code's limitations. (4) What is being done to remove these limitations. (5) How does the code compare with experimental observations and other code predictions. (6) What reference documents are available.

  18. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  19. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  20. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues; La production d'electricite d'origine nucleaire en France, dans le futur a long terme: Le cas des surgenerateurs: Les reacteurs nucleaires surgenerateurs: Les parametres physique et physico-chimiques, la thermodynamique associee des materiaux et de l'ingenierie mecanique: Nouveautes et options

    Energy Technology Data Exchange (ETDEWEB)

    Dautray, R. [Academie des sciences, 23, quai de Conti, 75270 Paris cedex 06 (France)

    2011-06-15

    The author gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the fifties. Neutron transport theory, thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, heat exchanges...) have now attained maturity, sufficient to implement sodium cooling circuits. However, the use of metallic sodium still raises certain severe questions in terms of safe handling and security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchangers) are undergoing in-depth research so as to last longer. The fuel cycle, notably the re-fabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts. (author)

  1. Ceramics for fusion reactors: The role of the lithium orthosilicate as breeder

    Energy Technology Data Exchange (ETDEWEB)

    Carella, Elisabetta, E-mail: elisabetta.carella@ciemat.es [National Laboratory for Magnetic Fusion, CIEMAT, Madrid (Spain); Hernandez, Teresa [National Laboratory for Magnetic Fusion, CIEMAT, Madrid (Spain)

    2012-11-15

    Lithium-based oxide ceramics are studied as breeder blanket materials for the controlled thermonuclear reactors (CTR). Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is one of the most promising candidates because of its lithium concentration (0.54 g/cm{sup 3}), its high melting temperature (1523 K) and its excellent tritium release behavior. It is reported that the diffusion of tritium is closely related to that of lithium, so it is possible to find an indirect measure of the trend of tritium studying the diffusivity of Li{sup +}. In the present work, the synthesis of the Li{sub 4}SiO{sub 4} is carried out by Spray drying followed by pyrolysis. The study of the Li{sup +} ion diffusion on the sintered bodies, is investigated by means of electrical conductivity measurements. The effect of the {gamma}-ray irradiation is evaluated by the impedance spectroscopy method (EIS) from room temperature to 1173 K. The results indicate that the sintesis process employed can produce Li{sub 4}SiO{sub 4} in the form of pebbles, finally the best ion species for the electrical conduction is the Li{sup +} and is shown that the g-irradiation to a dose of 5MGy, facilitate its mobility through the creation of defects, without change in its conduction process.

  2. Estimated recurrence frequencies for initiating accident categories associated with the Clinch River Breeder Reactor Plant design

    Energy Technology Data Exchange (ETDEWEB)

    Copus, E R

    1982-04-01

    Estimated recurrence frequencies for each of twenty-five generic LMFBR initiating accident categories were quantified using the Clinch River Breeder Reactor Plant (CRBRP) design. These estimates were obtained using simplified systems fault trees and functional event tree models from the Accident Delineation Study Phase I Final Report coupled with order-of-magnitude estimates for the initiator-dependent failure probabilities of the individual CRBRP engineered safety systems. Twelve distinct protected accident categories where SCRAM is assumed to be successful are estimated to occur at a combined rate of 10/sup -3/ times per year while thirteen unprotected accident categories in which SCRAM fails are estimated to occur at a combined rate on the order of 10/sup -5/ times per year. These estimates are thought to be representative despite the fact that human performance factors, maintenance and repair, as well as input common cause uncertainties, were not treated explicitly. The overall results indicate that for the CRBRP design no single accident category appears to be dominant, nor can any be totally eliminated from further investigation in the areas of accident phenomenology for in-core events and post-accident phenomenology for containment.

  3. Design of improved thermometer for the prototype fast breeder reactor MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Shimano, Kunio; Ito, Kenji; Tomobe, Katsuma [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Monju Construction Office, Tsuruga, Fukui (Japan)

    2002-12-01

    The thermometer design for the secondary coolant system was improved to prevent recurring failure of the thermometer well due to flow-induced vibration, the direct cause of the sodium leak incident of the prototype fast breeder reactor 'MONJU'. To satisfy the requirements of average temperature measurement, response time (within 20 seconds), avoidance and restraint of synchronized vibration, the insertion length of thermometer wells into the pipe was shortened to 110 mm for the response requirement and 60 mm for the no response requirement with a tapered shape. To simplify the installation, thermometer wells are mounted on the existing nozzles. To confirm the suitability of the design, analyses and experiments using the final design of the improved thermometer were performed. By analytical evaluation of flow-induced vibration and strength, the structural integrity was confirmed. Additionally, through flow-induced vibration experience, analyses of vibration characteristics confirmed the suitability. Furthermore, manufacture and welding of the thermometer wells on the existing nozzles were confirmed to be possible. (author)

  4. Performance characterization of geopolymer composites for hot sodium exposed sacrificial layer in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Haneefa, K. Mohammed, E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F.C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2013-12-15

    Highlights: • Performance evaluation of geopolymers subjected to hot liquid sodium is performed. • Apart from mechanical properties, micro-analytical techniques are used for material characterization. • The geopolymer composite showed comparatively lesser damage than conventional cement composites. • Geopolymer technology can emerge as a new choice for sacrificial layer in SCFBRs. - Abstract: A sacrificial layer of concrete is used in sodium cooled fast breeder reactors (SCFBRs) to mitigate thermo-chemical effect of accidentally spilled sodium at and above 550 °C on structural concrete. Performance of this layer is governed by thermo-chemical stability of the ingredients of sacrificial layer concrete. Concrete with limestone aggregate is generally used as a sacrificial layer. Conventional cement based systems exhibit instability in hot liquid sodium environment. Geo-polymer composites are well known to perform excellently at elevated temperatures compared to conventional cement systems. This paper discusses performance of such composites subjected to exposure of hot liquid sodium in air. The investigation includes comprehensive evaluation of various geo-polymer composites before any exposure, after heating to 550 °C in air, and after immersing in hot liquid sodium initially heated to 550 °C in air. Results from the current study indicate that hot liquid sodium produces less damage to geopolymer composites than to the existing conventional cement based system. Hence, the geopolymer technology has potential application in mitigating the degrading effects of sodium fires and can emerge as a new choice for sodium exposed sacrificial layer in SCFBRs.

  5. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail: maran@igcar.gov.in; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

    2015-10-15

    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  6. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  7. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Min; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Zhou, Guangming; Liu, Qianwen; Wang, Shuai; Lv, Zhongliang; Ye, Minyou

    2015-02-15

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%.

  8. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    Energy Technology Data Exchange (ETDEWEB)

    DeVan, J.H.

    1979-01-01

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF--BeF/sub 2/, Pb--Li alloys, and solid ceramic compounds such as Li/sub 2/O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies.

  9. Outlet plenum flow stratification studies for the Clinch River Breeder Reactor Plant

    Energy Technology Data Exchange (ETDEWEB)

    Novendstern, E.H.; Woods, M.D.; Andreychek, T.S.; Flannigan, L.J.; Carr, J.A.

    1976-01-01

    The transient temperature behavior during a simulated reactor trip was studied in a 1/3 sector, 0.55 scale model of the reactor outlet plenum of the CRBRP. Buoyancy effects were simulated using water and salt solution. Effects of variations in Richardson Number (ratio of buoyancy/inertia forces) and Froude Number (ratio of inertia/viscous forces) were evaluated. Effects of geometrical changes of a component in the outlet plenum called the Upper Internals Structure on transient temperature response were studied; both height of the structure and leakage under this component were experimentally varied. The test results confirm that flow stratification occurs in the outlet plenum following a reactor trip. The colder, denser fluid issuing from the core assemblies during the transient fills the lower portion of the plenum while the hotter fluid is trapped in the region above the outlet nozzles.

  10. Research and development in welding and hardfacing towards construction of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Albert, S.K.; Bhaduri, A.K. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2012-07-01

    India's 500MWe Prototype Fast Breeder Reactor (PFBR) is in advanced stage of construction at Kalpakkam and this reactor is expected to be commissioned in the year 2013. Extensive research and development activities in various fields like material development, welding, forming, non-destructive testing etc. were undertaken before the actual construction of the reactor began. Many of these activities are still continuing with the objectives of conducting functional tests, generating data, validating the design and meeting the various regulatory requirements. In welding, initial challenge was to develop indigenous welding consumables with a specification more stringent than that is given in most of the national and international standards. The welding consumable specified for 316LN austenitic stainless steel is E316-15M with strict control on delta ferrite content, toughness requirement after 750 C/100 h ageing to ensure adequate resistance to embrittlement during prolonged high temperature exposure and good slag detachability. This consumable was successfully developed in collaboration with Indian consumable manufacturers and is being used for fabrication of almost all PFBR components and piping made of 316LN stainless steel. Similarly, electrodes of welding of modified 9Cr-1Mo steel (material of construction for PFBR steam generator) with requirement of RTNDT requirement of ≤ -5 C was also developed indigenously. Extensive studies were also carried out on weldability of various austenitic stainless steels and modified 9Cr-1Mo steel used in PFBR. Hot cracking susceptibility of alloy D9 (15Cr-15Ni-2Mo-Ti alloy), the material chosen for fuel clab tube and fuel sub-assembly, and 316LN stainless steel was extensively studied using varestraint testing and Gleeble simulation. Results from these steels were used in developing welding procedures for various reactor components. Hydrogen assisted cracking susceptibility (HAC) of modified 9Cr-1Mo steel was studied using

  11. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  12. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. A review. Pt. I. Key areas

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions, Coimbatore, Tamilnadu (India)

    2013-11-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  13. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  14. Markovian reliability analysis under uncertainty with an application on the shutdown system of the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Papazoglou, I A; Gyftopoulos, E P

    1978-09-01

    A methodology for the assessment of the uncertainties about the reliability of nuclear reactor systems described by Markov models is developed, and the uncertainties about the probability of loss of coolable core geometry (LCG) of the Clinch River Breeder Reactor (CRBR) due to shutdown system failures, are assessed. Uncertainties are expressed by assuming the failure rates, the repair rates and all other input variables of reliability analysis as random variables, distributed according to known probability density functions (pdf). The pdf of the reliability is then calculated by the moment matching technique. Two methods have been employed for the determination of the moments of the reliability: the Monte Carlo simulation; and the Taylor-series expansion. These methods are adopted to Markovian problems and compared for accuracy and efficiency.

  15. The effect of water on tritium release behavior from solid breeder candidates

    Energy Technology Data Exchange (ETDEWEB)

    Suematsu, K.; Nishikawa, M.; Fukada, S.; Kinjyo, T.; Koyama, T.; Yamashita, N. [Graduate School of Engineering Science, Kyushu Univ., Fukuoka, 812-8581 (Japan)

    2008-07-15

    The authors have made a tritium release model to represent the release behavior of bred tritium from solid breeder materials using a series of studies. It has been observed that a large amount of adsorbed water and water produced by water formation reaction are released to the purge gas even though dry purge gas with hydrogen is introduced to solid breeder materials. According to our tritium release model, the presence of water in the purge gas and surface water on the material has a large effect on the tritium release behavior. In this study, the authors quantified the amount of adsorbed water and the capacity of the water formation reaction for various solid breeder materials (Li{sub 2}TiO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3}, LiAlO{sub 2}). The effect of surface water on the chemical form of tritium released from the LiAlO{sub 2} blanket is also discussed in this study. (authors)

  16. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  17. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  18. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  19. Numerical analysis of grid plate melting after a severe accident in a Fast-Breeder Reactor (FBR)

    Indian Academy of Sciences (India)

    A Jasmin Sudha; K Velusamy

    2013-12-01

    Fast breeder reactors (FBRs) are provided with redundant and diverse plant protection systems with a very low failure probability (<10-6/reactor year), making core disruptive accident (CDA), a beyond design basis event (BDBE). Nevertheless, safety analysis is carried out even for such events with a view to mitigate their consequences by providing engineered safeguards like the in-vessel core catcher. During a CDA, a significant fraction of the hot molten fuel moves downwards and gets relocated to the lower plate of grid plate. The ability of this plate to resist or delay relocation of core melt further has been investigated by developing appropriate mathematical models and translating them into a computer code HEATRAN-1. The core melt is a time dependent volumetric heat source because of the radioactive decay of the fission products which it contains. The code solves the nonlinear heat conduction equation including phase change. The analysis reveals that if the bottom of grid plate is considered to be adiabatic, melt-through of grid plate (i.e., melting of the entire thickness of the plate) occurs between 800 s and 1000 s depending upon the initial conditions. Knowledge of this time estimate is essential for defining the initial thermal load on the core catcher plate. If heat transfer from the bottom of grid plate to the underlying sodium is taken into account, then melt-through does not take place, but the temperature of grid plate is high enough to cause creep failure.

  20. Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Suzuki, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Ezato, Koichiro; Seki, Yohji; Yoshikawa, Akira; Tsuru, Daigo; Akiba, Masato [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan)

    2012-08-15

    The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

  1. Protection of broiler breeders by an inactivated combined water-in-oil-in-water viral vaccine.

    Science.gov (United States)

    Bidin, Z; Cajavec, S; Sladić, D; Ergotić, N; Cizelj, A; Pokrić, B

    1998-01-01

    A four-component vaccine, prepared by combining the single vaccines, contains subunits of Newcastle disease and infectious bronchitis viruses, as well as whole inactivated infectious bursal disease and egg drop syndrome viruses. The vaccine is prepared in the form of a low-viscosity water-in-oil-in-water emulsion with low mineral oil content. Heavy breeders were vaccinated at the age of 20 weeks by intramuscular administration of 0.5 ml vaccine/bird in an experiment carried out under field conditions, involving 5000 female and 450 male parents. The birds had previously been vaccinated with live vaccines according to an obligatory field vaccination programme. Vaccination with the WOWE vaccine near the point of lay elicited serological responses protecting both the parents and their progeny. Each of the antigens administered in the four-component vaccine was as effective as the respective single component vaccine. The mortality, recorded during the 31-week experimental period, was 6.2%. Mortality and morbidity were not triggered by viruses against which vaccination was carried out. Egg production was not affected by the vaccination and was 170.2 eggs per hen during the 28-week production period.

  2. Considerations on a critical experiment program for a large fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-09-15

    The design studies for large LMFBR in Japan are being continued by PNC and Electric Companies. Both of them have adopted loop-type and 1000 MWe class reactor as the reference design, but main parameters of the core have not yet been fixed. The main core parameters of the present design are shown. Comparing the geometrical properties with those in ZPPR-9 and 10, some of the degrees of mockup are not satisfactory. In addition, there is another difference between a reactor and a mockup critical assembly. Therefore extrapolation is important to apply the results of JUPITER to the core design.

  3. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  4. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A. (eds.)

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  5. Automated operator procedure prompting for startup of Experimental Breeder Reactor-2

    Energy Technology Data Exchange (ETDEWEB)

    Renshaw, A.W.; Ball, S.J.; Ford, C.E.

    1990-11-01

    This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.

  6. Linearized model for the hydrodynamic stability investigation of molten fuel jets into the coolant of a Liquid Metal Fast Breeder Reactor (LMFBR)

    Science.gov (United States)

    Hartel, K.

    1986-02-01

    The hydrodynamic stability of liquid jets in a liquid continuum, both characterized by low viscosity was analyzed. A linearized mathematical model was developed. This model enables the length necessary for fragmentation of a vertical, symmetric jet of molten fuel by hydraulic forces in the coolant of a liquid metal fast breeder reactor to be evaluated. On the basis of this model the FRAG code for numerical calculation of the hydrodynamic fragmentation mechanism was developed.

  7. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    Energy Technology Data Exchange (ETDEWEB)

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  8. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    Science.gov (United States)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  9. Evaluation on the heat removal capacity of the first wall for water cooled breeder blanket of CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng, E-mail: jiangkecheng@ipp.ac.cn; Cheng, Xiaoman; Chen, Lei; Huang, Kai; Ma, Xuebin; Liu, Songlin

    2016-02-15

    Highlights: • Heat removal capacity of the FW is evaluated under BWR, PWR and He coolant inlet conditions. • Heat transfer property of the gas–liquid two phase and the two boiling crises are analyzed. • Heat removal capacity of water is larger than helium coolant. - Abstract: The water cooled ceramic breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). As an important component of the blanket, the FW should satisfy with the thermal requirements in any case. In this paper, three parameters including the heat removal capacity, coolant pressure drop as well as the temperature rise of the FW were investigated under different coolant velocity and heat flux from the plasma. Using the same first wall structure, two main water cooled schemes including Boiling Water Reactor (BWR, 7 MPa pressure and 265 °C temperature inlet) and Pressurized Water Reactor (PWR, 15 MPa pressure and 285 °C temperature inlet) conditions are discussed in the thermal hydraulic calculation. For further research, the thermal hydraulic characteristics of using helium as coolant (8 MPa pressure, 300 °C temperature inlet) are also explored to provide CFETR blanket design with more useful data supports. Without regard to the outlet coolant condition requirements of the blanket, the results indicate that the ultimate heat flux that the FW can resist is 2.2 MW/m{sup 2} at velocity of 5 m/s for BWR, 2.0 MW/m{sup 2} at velocity of 5 m/s for PWR and 0.87 MW/m{sup 2} for helium at velocity 100 m/s under the chosen operation condition. The detrimental departure from nucleate boiling (DNB) crisis would occur at the velocity of 1 m/s under the heat flux of 3 MW/m{sup 2} and dry out crisis appears at the velocity of less than 0.2 m/s with the heat flux of more than 1 MW/m{sup 2} for BWR. The further blanket/FW optimization design is provided with more useful data references according to the abundant calculation results.

  10. Kinematic dynamo action in a network of screw motions; application to the core of a fast breeder reactor

    Science.gov (United States)

    Plunian, F.; Marty, P.; Alemany, A.

    1999-03-01

    Most of the studies concerning the dynamo effect are motivated by astrophysical and geophysical applications. The dynamo effect is also the subject of some experimental studies in fast breeder reactors (FBR) for they contain liquid sodium in motion with magnetic Reynolds numbers larger than unity. In this paper, we are concerned with the flow of sodium inside the core of an FBR, characterized by a strong helicity. The sodium in the core flows through a network of vertical cylinders. In each cylinder assembly, the flow can be approximated by a smooth upwards helical motion with no-slip conditions at the boundary. As the core contains a large number of assemblies, the global flow is considered to be two-dimensionally periodic. We investigate the self-excitation of a two-dimensionally periodic magnetic field using an instability analysis of the induction equation which leads to an eigenvalue problem. Advantage is taken of the flow symmetries to reduce the size of the problem. The growth rate of the magnetic field is found as a function of the flow pitch, the magnetic Reynolds number (Rm) and the vertical magnetic wavenumber (k). An [alpha]-effect is shown to operate for moderate values of Rm, supporting a mean magnetic field. The large-Rm limit is investigated numerically. It is found that [alpha]=O(Rm[minus sign]2/3), which can be explained through appropriate dynamo mechanisms. Either a smooth Ponomarenko or a Roberts type of dynamo is operating in each periodic cell, depending on k. The standard power regime of an industrial FPBR is found to be subcritical.

  11. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  12. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  13. Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2016-05-15

    Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

  14. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This report includes engineering memorandums, drawings, key feature descriptions, and other data. Some of the reports, such as manufacturability and some stress analysis, were done by consultants for Byron Jackson. Review of this report indicates that the design is feasible. The pump can be manufactured to system and specification requirements. The overall length and weight of some pieces will require special consideration, but is within the scope of equipment and technology available today. The fabricated parts are large and heavy, but can be manufactured and machined. Only the high temperature is unique to this size, since previous sodium pumps were smaller. Nondestructive tests as required by the Code are described and are feasible. The performance test of the prototype has been studied thoroughly. It is feasible for a cold water test. There are some problem areas. However, all of them can be solved. Development needs include building and testing a small scale model.

  15. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. Pt. II. R and D necessities and development across the world. A review

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions Coimbatore, Tamilnadu (India)

    2013-12-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R and D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I. (orig.)

  16. Localized corrosion studies on materials proposed for a safety-grade sodium-to- air decay-heat removal system for fast breeder reactors

    Science.gov (United States)

    Kamachi Mudali, U.; Khatak, H. S.; Dayal, R. K.; Gnanamoorthy, J. B.

    1993-02-01

    The present investigation was carried out to assess the localized corrosion resistance of materials proposed for the construction of the safety-grade sodium-to-air decay-heat removal system for fast breeder reactors. The materials, such as Alloy 800,9Cr-lMo steel, and type 316LN stainless steel, in different microstructural conditions were assessed for pitting and stress-corrosion cracking resistances in a chloride medium. The results indicated that 9Cr-lMo steel in the normalized and tempered condition can be considered for the above application from the standpoint of corrosion resistance.

  17. OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels.

  18. Heavy Water Reactor; Reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Yu, St.; HOpwood, J.; Meneley, D. [Energie Atomique du Canada (Canada)

    2000-04-01

    This document deals with the Heavy Water Reactor (HWR) technology and especially the Candu (Canada Deuterium Uranium) reactor. This reactors type offers many advantages that promote them for the future. General concepts, a description of the Candu nuclear power plants, the safety systems, the fuel cycle and economical and environmental aspects are included. (A.L.B.)

  19. Analysis of unprotected transients with control and safety rod drive mechanism expansion feedback in a medium sized oxide fuelled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sathiyasheela, T., E-mail: sheela@igcar.gov.in; Natesan, K.; Srinivasan, G.S.; Devan, K.; Puthiyavinayagam, P.

    2015-09-15

    Highlights: • Possibilities of enhancing safety under ULOF and UTOP accidents. • CSRDM expansion feedbacks under unprotected transients. • CSRDM expansion feedback enhances the safety of fast reactors. • CSRDM expansion feedbacks ensuring enough time for initiating safety actions. - Abstract: Possibilities of enhancing core safety under unprotected loss of flow (ULOF) and unprotected transient over power (UTOP) accidents with control and safety rod drive mechanism (CSRDM) expansion feedbacks are explored in a medium sized oxide fuelled fast breeder reactor. This feedback is expected to take the reactor to a safe shutdown under ULOF and to an another steady state under UTOP where there is no significant fuel melting. Under ULOF, with CSRDM feedback net reactivity was maintained negative throughout the transient (up to 2000 s) and the power dropped to a level of heat removal capacity of decay heat removal system based on natural circulation. Similarly, under UTOP with the above feedback reactor power goes to a lower peak value. The fuel temperature is just touching the melting temperature and the melt fraction does not cross 5%. With CSRDM expansion feedbacks both ULOF and UTOP transients prolong beyond 2000 s. It ensures, availability of time for initiating any safety actions against the transients, and thus it helps to preclude core disruptive accidents (CDA) in a medium sized oxide fuelled reactors.Classification: L. safety and risk analysis.

  20. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  1. Optimization of the breeder zone cooling tubes of the DEMO Water-Cooled Lithium Lead breeding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A.; Arena, P.; Bongiovì, G. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Chiovaro, P., E-mail: pierluigi.chiovaro@unipa.it [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy); Del Nevo, A. [ENEA Brasimone, Camugnano, BO (Italy); Forte, R. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo, Viale delle Scienze, Palermo (Italy)

    2016-11-01

    Highlights: • Determination of an optimal configuration for the breeder zone cooling tubes. • Attention has been focused on the toroidal–radial breeder zone cooling tubes lay out. • A theoretical-computational approach based on the Finite Element Method (FEM) has been followed, adopting a qualified commercial FEM code. • Five different configurations have been investigated to optimize the breeder zone cooling tubes arrangement fulfilling all the rules prescribed by safety codes. - Abstract: The determination of an optimal configuration for the breeder zone (BZ) cooling tubes is one of the most important issues in the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket R&D activities, since BZ cooling tubes spatial distribution should ensure an efficient heat power removal from the breeder, avoiding hotspots occurrence in the thermal field. Within the framework of R&D activities supported by the HORIZON 2020 EUROfusion Consortium action on the DEMO WCLL breeding blanket design, a campaign of parametric analyses has been launched at the Department of Energy, Information Engineering and Mathematical Models of the University of Palermo (DEIM), in close cooperation with ENEA-Brasimone, in order to assess the potential influence of BZ cooling tubes number on the thermal performances of the DEMO WCLL outboard breeding blanket equatorial module under the nominal steady state operative conditions envisaged for it, optimizing their geometric configuration and taking also into account that a large number of cooling pipes can deteriorate the tritium breeding performances of the module. In particular, attention has been focused on the toroidal-radial option for the BZ tube bundles lay-out and a parametric study has been carried out taking into account different tube bundles arrangement within the module. The study has been carried out following a numerical approach, based on the finite element method (FEM), and adopting a qualified commercial FEM code. Results

  2. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  3. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  4. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  5. Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

  6. Summary of the 3rd workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

    2000-06-01

    The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

  7. Summary of the 4th workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  8. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    OpenAIRE

    Matjaž Leskovar; Mitja Uršič

    2016-01-01

    A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...

  9. Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

    Energy Technology Data Exchange (ETDEWEB)

    Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-28

    Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

  10. Control rod calibration methods for fast breeder reactors applied to Phenix; Les methodes d'etalonnage des barres de commande des reacteurs a neutrons rapides application a Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Lecourt, G

    1998-06-18

    The control and the emergency shutdown of a fast breeder reactor depends essentially on control rods. For this reason, it is imperative to know exactly how much anti reactivity is introduced with the rods in the reactor core. Different methods have been compared in order to see if they are compatible with Phenix reactor. Their limits have been studied. The shadow and anti shadow effects that can the rods make one to the other and then their effective weight of the rods screen have been clarified. (N.C.)

  11. Numerical studies on the heat transfer and friction characteristics of the first wall inserted with the screw blade for water cooled ceramic breeder blanket of CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Kecheng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230037 (China); Ma, Xuebin; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2016-03-15

    Highlights: • Enhanced heat transfer and friction characteristics of the FW inserted with screw blade is investigated. • The screw blade structure optimization was done on the screw pitch and diameter. • Decreasing screw pitch and increasing screw diameter could further enhance heat transfer accompanied with increasing flow resistance. • Evaluate the overall enhanced heat performance by using the PEC value. - Abstract: The Water Cooled Ceramic Breeder (WCCB) blanket based on Pressurized Water Reactor (PWR) condition is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). The first wall (FW) which plays an important part in the blanket design must remove the high heat flux radiated from plasma and nuclear heat deposition on the structure in any operating conditions. In this paper, the characteristics of enhanced heat transfer and friction for the FW with the inserted screw blade are studied by the numerical method. After the comparison between the numerical and experimental results, the standard k–ε turbulent model is selected to do the numerical calculation. The numerical results show that the peak temperature of RAFM steel could be reduced by decreasing screw pitch or increasing screw diameter, while accompanied with ascending flow resistance. Besides, among all of the chosen calculation cases compared with the smooth channel, the maximum value of temperature reduction is 10 °C under the conditions of heat flux of 0.5 MW/m{sup 2} as well as screw pitch of 18 mm and screw diameter of 6 mm. The maximum increment ratio of the friction factor is 257% under the conditions of screw pitch of 10 mm and screw diameter of 4 mm. Furthermore, screw blade of 74 mm pitch and 4 mm diameter presents the highest overall performance evaluation criterion (PEC) value of 0.93 under Reynolds number of 270 000 conditions, and shows the best overall heat transfer enhancement performance.

  12. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) neat transport system dynamics and steam generator control

    Science.gov (United States)

    Brukx, J. F. L. M.

    1982-06-01

    Loop type LMFBR heat transport system dynamics after reactor shutdown and during subsequent decay heat removal are considered with emphasis on steam generator dynamics including the development and evaluation of various post-scram steam generator control systems, and natural circulation of the sodium coolant, including the influence of superimposed free convection on forced convection heat transfer and pressure drop. The normal operating and decay heat removal functions of the overall heat transport system are described.

  13. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Science.gov (United States)

    Ioltukhovskiy, A. G.; Leonteva-Smirnova, M. V.; Solonin, M. I.; Chernov, V. M.; Golovanov, V. N.; Shamardin, V. K.; Bulanova, T. M.; Povstyanko, A. V.; Fedoseev, A. E.

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a δ-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 °C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 °C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  14. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ioltukhovskiy, A.G. E-mail: iral@bochvar.ru; Leonteva-Smirnova, M.V.; Solonin, M.I.; Chernov, V.M.; Golovanov, V.N.; Shamardin, V.K.; Bulanova, T.M.; Povstyanko, A.V.; Fedoseev, A.E

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a {delta}-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 deg. C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 deg. C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  15. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  16. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Matjaž Leskovar

    2016-02-01

    Full Text Available A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

  17. LMFBR (LIQUID METAL FAST BREEDER REACTOR) READTION RATE AND DOSIMETRY 3RD QUARTERLY PROGRESS REPORT DECEMBER 1971 JANUARY FEBRUARY 1972

    Energy Technology Data Exchange (ETDEWEB)

    MCELROY WN

    1972-03-01

    This report was compiled at the Hanford Engineering Development Laboratory operated by Westinghouse Hanford Company, a subsidiary of Westinghouse Electric Corporation, for the United States Atomic Energy Commission, Division of Reactor Development and Technology, under Contract No. AT (45-1) 2170. It describes technical progress made in the Interlaboratory LMFBR Reaction Rate Program during the reporting period. The Interlaboratory LMFBR Reaction Rate (ILRR) program has been established by USAEC/RDT to develop a capability to accurately measure neutron-induced reaction rates for LMFBR fuels and materials development programs. The initial goal for the principal fission reactions, {sup 235}U, {sup 238}U, and {sup 239}Pu, is an accuracy to within {+-}5 at the 95% confidence level. Accurate measurement of other fission and non-fission reactions will be required, but to a lesser accuracy, between {+-}5 to 10% at the 95% confidence level. A secondary program objective is improvement in knowledge of the nuclear parameters involved in fuels and materials dosimetry measurements of neutron flux, spectra, fluence, and burnup. These accuracy goals for the ILRR program are severe; measurements of fast-neutron-induced reaction rates have not been rapidly moving toward this level of precision. Using a number of techniques in well established neutron environments of current interest for fast reactor development and critically evaluating the results will help establish existing levels of accuracy and indicate the scale of effort required for improvement. To accomplish the objectives of this program, reliable and documented experimental values of reaction rates and ratios will be determined for various well established and permanent neutron fields. The Coupled Fast Reactivity Measurement Facility (CFRMF) at Aerojet Nuclear Company (ANC) is the first neutron field being studied because of the similarity of its spectrum to that of a fast reactor and the range and reproducibility of

  18. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  19. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  20. Role of energetic mixed-oxide-fuel-sodium thermal interactions in liquid metal fast breeder reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Fauske, H.K.

    1976-01-01

    Based upon analysis, numerous experiments and examination of all known occurrences of large-mass vapor explosions, the following general behavior principle has emerged: Mixing of large quantities of a hot and cold liquid, a necessary condition for developing sustained pressures and large damage potential from thermal interaction, requires spontaneous nucleation upon contact. Since the contact temperature for the mixed-oxide-fuel-sodium system is well below the spontaneous-nucleation temperature for liquid sodium, the current interesting controversy regarding spontaneous nucleation and its role in the vapor-explosion mechanism itself is largely irrelevant for this system. Therefore, current practice is to use the pressure-volume curve determined by the expanding fuel vapor following a postulated hydrodynamic disassembly (which generally results from considering a number of unrealistic physical processes to occur) for safety evaluation. It follows that for reactors like FFTF and CRBR, the extremely unlikely event of a core meltdown is predicted to occur safely, with essentially no energetics involved.

  1. 快堆钠回路水锤程序开发与应用%Waterhammer Program Development and Application for Fast Breeder Reactor's Sodium Circus

    Institute of Scientific and Technical Information of China (English)

    文静; 栾霖; 金德圭; 陆道纲; 汤荣铭

    2001-01-01

    研究开发了快堆钠回路水锤分析专用程序WHA。该程序在一维特征线法(MOC)传统的压力波传播数学模型中补充了钠腔-气腔外边界模型,并采用气泡离散模型模拟低压液柱分离中的蒸汽穴的生成与溃灭。程序用FORTRAN90语言对快堆实验钠回路ESPRESSO中由于阀门的快速开启与关闭引起的压力波传播进行了分析计算。计算结果表明:将钠腔-气腔引入水锤压力波传播的数学模型进行程序计算的结果是合理的。%Based on one-dimensional method of characteristics(MOC), anumerical model of pressure-wave progation is presented in the paper. A special code is programmed to analyze and calculate waterhammer resulted from rapid opening or closing of valve in the experimental sodium circus of fast breeder reactor(FBR). In the model, a new outer boundary condition, sodium-cavity is included. Model of bubble's discrete distribution is adopted to simulate generation and collapse of the bubble with the pressure's decreasing and increasing. The results demonstrate that the model of pressure-wave progation is valid.

  2. Study on water cooled high conversion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of study on advanced reactors for the future, conceptual design of high conversion water cooled reactors is being studied, aiming at the contribution to nuclear fuel cycle by the LWR technology, since the utilization of LWRs will extend over a long period of time . We are studying on the reactor core concepts for BWR and PWR reactor systems. As for BWR system, three types of reactor cores are investigating for three different design goals; long operation period, high conversion ratio and high applicability for the existing BWR system. In all the cases, we have obtained a fair prospect of a large core concept with a capacity of 1,000 MWe class having negative void reactivity coefficient. This study is a part of JAERI-JAPCO (Japan Atomic Power Company) cooperative studies. Various kinds of conceptual designs will be created until the end of FY 1999. The designs will be checked and reviewed at that time, then experimental studies on the realization of the concepts will start with further design works from FY 2000. (author)

  3. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    Science.gov (United States)

    Treshow, M.

    1963-04-30

    This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

  4. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  5. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  6. Advanced light water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Giedraityte, Zivile [Helsinki University of Technology, Otaranta 8D-84, 02150 Espoo (Finland)

    2008-07-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  7. Helium-cooled molten-salt fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; Devan, J.H.

    1984-12-01

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF/sub 2/ + ThF/sub 4/) is circulated through the blanket and to the on-line processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of /sup 233/U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the /sup 233/U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned.

  8. Field study on evaluation of the efficacy and usability of two disinfectants for drinking water treatment at small cattle breeders and dairy cattle farms.

    Science.gov (United States)

    Mohammed, Asmaa N

    2016-03-01

    The hygienic quality of drinking water for cattle originated from different sources together with the efficacy and usability of two types of disinfectants against waterborne pathogens were assessed for small cattle breeders and dairy cattle farms. A total of 120 drinking water samples were collected from water troughs representing three different water sources commonly used for cattle drinking (tap, underground and surface water; n = 65, 25, and 30, respectively). Collected samples were cultured for isolation and identification of pathogenic bacteria using serological techniques and PCR. The bactericidal efficacy of the disinfectants, sodium dichloroisocyanurate (NaDCC) and hydrogen peroxide (H2O2) 50%, at different concentrations were evaluated by the determination of total viable and coliform counts of water prior and postwater treatment. In small cattle breeders, Escherichia coli was the most prevalent bacterial isolates from surface water (56.7%) followed by Staphylococcus aureus (36.7%), Salmonella spp. (26.7%), Streptococcus faecalis (23.3%), Shigella flexneri (16.7%), Proteus spp. (16.7%), and Klebsiella pneumonae (10.0 %) at X(2) = 9, P ≤ 0.01. Prior to the use of disinfectants, the averages of total bacterial and coliform counts were the highest in surface water (3.56 × 10(7), 240.0, and 38.0 CFU/100 ml, respectively). It has been found that hydrogen peroxide 50% at a concentration of 35 mg/l had a lethal effect (100 %) on indicator microorganisms compared with NaDCC at concentration of 2 mg/l. In conclusion, the higher bacterial contaminants in drinking water were found in surface water followed by tap water, particularly for small cattle breeders. Therefore, the usage of more hygienic water troughs with their regular treatment by hydrogen peroxide 50% at concentration of 35 mg/l is highly recommended to control waterborne bacteria and consequently improve and maintain the animal health.

  9. Thermal breeder fuel enrichment zoning

    Science.gov (United States)

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  10. Characterization and improvement of water compatibility of {gamma}-LiAlO{sub 2} ceramic breeders

    Energy Technology Data Exchange (ETDEWEB)

    Lin Jiu [Shanghai Institute of Ceramics, Chinese Academy of Sciences, 1295 DingXi Road, Shanghai 200050 (China); Wen Zhaoyin, E-mail: zywen@mail.sic.ac.cn [Shanghai Institute of Ceramics, Chinese Academy of Sciences, 1295 DingXi Road, Shanghai 200050 (China); Xu Xiaogang; Li Ning; Song Shufeng [Shanghai Institute of Ceramics, Chinese Academy of Sciences, 1295 DingXi Road, Shanghai 200050 (China)

    2010-12-15

    The influence of water treatment at 473 K on the phase and morphology of the {gamma}-LiAlO{sub 2} ceramics was studied. It was implied that a AlO(OH) layer was formed on the surface of the {gamma}-LiAlO{sub 2} ceramics, at the same time LiAl{sub 2}(OH){sub 7}.xH{sub 2}O was obtained in the water. The process of the interaction between {gamma}-LiAlO{sub 2} ceramics and water was deduced. The attempts to improve the water compatibility of {gamma}-LiAlO{sub 2} ceramics were made. It was confirmed that the surface modification by LiAl{sub 5}O{sub 8} decreased the surface alkalinity and restrained the hydrolysis of the {gamma}-LiAlO{sub 2} ceramics in water due to the low hydrolysis of the lithium deficient phase LiAl{sub 5}O{sub 8}. The experimental results revealed that the introduction of the lithium deficient phase, LiAl{sub 5}O{sub 8}, effectively improved the water compatibility of the {gamma}-LiAlO{sub 2} ceramics.

  11. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

    2001-07-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  12. A method for improvement of safety features of large fast breeder reactors. Numerical simulation of unprotected loss-of-flow accident in an LMFBR equipped with gas-expansion modules

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Masayoshi [Hitachi Engineering Co. Ltd., Ibaraki (Japan); Murakami, Tomoko; Kawashima, Katsuyuki; Watari, Yoshio; Nakao, Noboru; Miura, Masanori

    1995-04-01

    Numerical simulation of an unprotected loss-of-flow (ULOF) accident has been performed for a large liquid-metal-cooled fast breeder reactor (LMFBR) equipped with gas expansion modules (GEMs) in the radial periphery of the reactor core. The effectiveness of the GEMs in small fast reactors was demonstrated already in the passive safety testing in the Fast Flux Test Facility. According to neutronic calculations based on the transport theory, even in large reactors of electrical power 600 to 1,300 MW, the reactivity worth of GEMs, which replace one layer of radial blanket fuel subassemblies, ranges from -1.9$ to -1.4$, depending on the size of the core. A simulation of ULOF transient was performed with a 5.5s flow-halving time in a 600 MWe LMFBR equipped with GEMs of -1.9$ reactivity worth. The result showed that, if 10% of the rated core coolant flow by pony motors was available following the main pump coastdown, the GEM reactivity alone could bring the reactor subcritical and the predicted maximum coolant temperature was substantially lower than the sodium boiling point. The reactivity worth calculations, a modeling of gas expansion behavior, and ULOF simulation together with needs of further development for the GEM application are described. (author).

  13. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  14. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  15. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  16. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  17. Environmentally assisted cracking in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  18. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  19. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  20. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  1. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  2. Computer simulation of the NASA water vapor electrolysis reactor

    Science.gov (United States)

    Bloom, A. M.

    1974-01-01

    The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

  3. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity. The feedback reactivity is a function of the operating conditions of the system, and cannot be directly measured. However, power measurements are relatively easy to obtain from the nuclear instrumentation and process computer, and are used in conjunction with a reduced order model to estimate the gain of the thermal-hydraulics feedback using an Extended Kalman Filter. The reduced order model is obtained by estimating the thermal-hydraulic transfer function from the frequency-domain BWR code LAPUR, and the stability properties are evaluated based on the pair of complex conjugate eigenvalues. Because of the recursive nature of the Kalman Filter, an estimate of the decay ratio is generated every sampling time, allowing continuous estimation of the stability parameters. A test platform based on a nuclear-coupled boiling channel is developed to validate the capability of the BWR stability monitoring methodology. The thermal-hydraulics for the boiling channel is modeled and coupled with neutron kinetics to analyze the non-linear dynamics of the closed-loop system. The model uses point kinetics to study core-wide oscillations, and normalized modal kinetics are introduced to study out-of-phase oscillations. The coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate noisy power time series. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions

  4. Environmentally assisted cracking in light water reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  5. Fabrication, properties, and tritium recovery from solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E. (Argonne National Lab., IL (USA)); Kondo, T. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Roux, N. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)); Tanaka, S. (Tokyo Univ. (Japan)); Vollath, D. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  6. Shielding designs for pressurized water reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Forestier, J.; Vergnaud, T.

    1986-07-01

    The efforts made by Electricite de France to reduce exposure from the two-component neutron-gamma radiation fields inside the pressurized water reactor (PWR) building are described. Most of the attention had been focused on the problem of neutron exposure relative to the problem of achieving a highly efficient confinement within the reactor cavity and the state of the art of personnel neutron dosimetry. A description of the general neutron calculation scheme that links the characteristics of the neutron fields escaping from the reactor vessel to the dose equivalent rate cartographies inside the reactor building is provided.

  7. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  8. Measuring motivation for appetitive behaviour: food-restricted broiler breeder chickens cross a water barrier to forage in an area of wood shavings without food.

    Directory of Open Access Journals (Sweden)

    Laura M Dixon

    Full Text Available Broiler breeders (parents of meat chickens are selected for fast growth and become obese if fed ad libitum. To avoid this and maintain good health and reproductive ability, they are feed restricted to about 1/3 of what they would eat ad libitum. As a result, they experience chronic hunger and exhibit abnormal behaviour patterns that may indicate stress and frustration. One approach to measuring hunger is to observe how much birds will work, such as pecking a key, for access to more or different types of food. However, the sight, smell, and feedback from consumption of the feed reward changes the context and may artificially raise feeding motivation. To avoid this, we tested broiler breeders in an apparatus in which they could work for access to a wooden platform covered in wood shavings by crossing a water runway which increased in length and depth in 8 successive tests. In the wood shavings area, they could perform exploratory and foraging behaviour (the appetitive phase of feeding but were never rewarded with feed. Sixty birds were divided into three feed quantity treatments: commercial restriction (R, and twice (2R or three times (3R this amount. Overall, birds fed R worked harder to reach the wood shavings area (reached it in a larger number of tests than 2R and 3R birds (P2R>3R. This indicates that restricted-fed birds were hungry and willing to work for the opportunity to forage even though food was never provided, suggesting that their motivation to perform the appetitive component of feeding behaviour (foraging/food searching was sufficient to sustain their response. Thus food restriction in broiler breeders is a welfare concern. However these methods could be used to test alternative feeding regimes to attempt to find ways of alleviating hunger while still maintaining healthy growth and reproduction in these birds.

  9. Measuring motivation for appetitive behaviour: food-restricted broiler breeder chickens cross a water barrier to forage in an area of wood shavings without food.

    Science.gov (United States)

    Dixon, Laura M; Brocklehurst, Sarah; Sandilands, Vicky; Bateson, Melissa; Tolkamp, Bert J; D'Eath, Rick B

    2014-01-01

    Broiler breeders (parents of meat chickens) are selected for fast growth and become obese if fed ad libitum. To avoid this and maintain good health and reproductive ability, they are feed restricted to about 1/3 of what they would eat ad libitum. As a result, they experience chronic hunger and exhibit abnormal behaviour patterns that may indicate stress and frustration. One approach to measuring hunger is to observe how much birds will work, such as pecking a key, for access to more or different types of food. However, the sight, smell, and feedback from consumption of the feed reward changes the context and may artificially raise feeding motivation. To avoid this, we tested broiler breeders in an apparatus in which they could work for access to a wooden platform covered in wood shavings by crossing a water runway which increased in length and depth in 8 successive tests. In the wood shavings area, they could perform exploratory and foraging behaviour (the appetitive phase of feeding) but were never rewarded with feed. Sixty birds were divided into three feed quantity treatments: commercial restriction (R), and twice (2R) or three times (3R) this amount. Overall, birds fed R worked harder to reach the wood shavings area (reached it in a larger number of tests) than 2R and 3R birds (P2R>3R). This indicates that restricted-fed birds were hungry and willing to work for the opportunity to forage even though food was never provided, suggesting that their motivation to perform the appetitive component of feeding behaviour (foraging/food searching) was sufficient to sustain their response. Thus food restriction in broiler breeders is a welfare concern. However these methods could be used to test alternative feeding regimes to attempt to find ways of alleviating hunger while still maintaining healthy growth and reproduction in these birds.

  10. High-fibre pelleted rations decrease water intake but do not improve physiological indexes of welfare in food-restricted female broiler breeders.

    Science.gov (United States)

    Hocking, P M

    2006-02-01

    1. A 3x2 factorial experiment was conducted with three diets and two lines of broiler breeder females to evaluate the contribution of low-energy rations for improving the welfare of feed-restricted birds during rearing. Experimental diets were fed from 6 to 16 weeks of age and were created by diluting a conventional grower (Control) ration containing 11.0 MJ ME/kg with 200 (8.8 MJ ME/kg) or 400 (6.6 MJ ME/kg) g oat hulls/kg using Optimoist to facilitate the pelleting process. Welfare was assessed by changes in behaviour and physiological variables at 8, 12 and 16 weeks of age. Birds were fed restricted quantities of feed to meet recommended body weight targets. 2. There was a decrease in the proportion of observations of drinking and an increase of preening in birds fed on the two experimental diets compared with the control. There was a linear decrease in litter moisture and the number of litter changes with increasing diet dilution, and water intake at 12 weeks was higher in the control than in the two experimental diets. There were no changes in physiological indexes of welfare (heterophil-lymphocyte ratio, plasma corticosterone and antibody responses) associated with the dietary treatments. 3. There were no important differences in the growth, behaviour or physiological responses to dietary treatment between the two lines of broiler breeders. Changes with age were similar to those reported in other experiments. 4. It was concluded that low-energy pelleted diets would improve litter conditions but not improve indexes of welfare in feed-restricted broiler breeders.

  11. Capital Cost: Pressurized Water Reactor Plant Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate.

  12. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  13. Tar water digestion in an upflow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

    Skibsted Mogensen, A.; Angelidaki, I.; Schmidt, J.E.; Ahring, B.K. [Technical Univ., Dept. of Environmental Science and Engineering, Lyngby (Denmark)

    1998-08-01

    The water from the gasification and wet oxidised tar water has been digested anaerobically in UASB reactors and were digested in respectively 10 and 50% in batches. Though the tar water show inhibition at very low concentrations to aerobic microorganisms, the granular sludge used in UASB reactors degrades tar water in concentrations that reveal total inhibition of e.g. bacteria conducting the nitrification process. The value of waste waters are determined, showing that the tar water produces more biogas in the anaerobic digestion. A wide range of xenobiotics, especially phenolic compounds can be transformed in the anaerobic digestion process. Seven phenolic are followed in batch experiments and UASB reactor experiments, and their particular fate in the anaerobic systems embody large differences in the transformation pattern. (au) 24 refs.

  14. Optimization of uranium use in light water reactors; Optimisation de l'utilisation des ressources dans les reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Greneche, D.; Lecomte, M. [AREVA NP, Tour AREVA 92 - Paris La Defense (France)

    2010-07-01

    Light water reactors are expected to produce most part of nuclear power for this century before giving way to breeder reactors so uranium resources have to be dealt with diligently. 5 ways to minimize the consumption of uranium in the fuel cycle are considered: 1) the optimization of the enrichment in order to reduce uranium tails; 2) a better use of uranium in the reactor through either higher burnups or a reduction of sterile neutron captures; 3) to get a better LWR's thermodynamical yield; 4) to improve plutonium and uranium recycling; 5) to use the stocks of existing fissile materials like depleted uranium, spent fuels, military plutonium; and 6) the development of the thorium cycle as a complement. (A.C.)

  15. Antineutrino monitoring for the Iranian heavy water reactor

    CERN Document Server

    Christensen, Eric; Jaffke, Patrick; Shea, Thomas

    2014-01-01

    In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

  16. Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium–Plutonium Carbide based fuel and 300 MWt Power

    Science.gov (United States)

    Clief Pattipawaej, Sandro; Su’ud, Zaki

    2017-01-01

    A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

  17. Light Water Reactor Sustainability Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  18. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  19. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  20. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  1. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  2. A comparative study of kinetics of nuclear reactors

    Directory of Open Access Journals (Sweden)

    Obaidurrahman Khalilurrahman

    2009-01-01

    Full Text Available The paper deals with the study of reactivity initiated transients to investigate major differences in the kinetics behavior of various reactor systems under different operating conditions. The article also states guidelines to determine the safety limits on reactivity insertion rates. Three systems, light water reactors (pressurized water reactors, heavy water reactors (pressurized heavy water reactors, and fast breeder reactors are considered for the sake of analysis. The upper safe limits for reactivity insertion rate in these reactor systems are determined. The analyses of transients are performed by a point kinetics computer code, PKOK. A simple but accurate method for accounting total reactivity feedback in kinetics calculations is suggested and used. Parameters governing the kinetics behavior of the core are studied under different core states. A few guidelines are discussed to project the possible kinetics trends in the next generation reactors.

  3. Corrosion and Corrosion Control in Light Water Reactors

    Science.gov (United States)

    Gordon, Barry M.

    2013-08-01

    Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

  4. Core management of the prototype heavy water reactor FUGEN

    Energy Technology Data Exchange (ETDEWEB)

    Deshimaru, Takehide; Furubayashi, Toshiyuki; Matsumoto, Mitsuo (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

    1982-12-01

    In this paper, the core management which has been implemented so far for the prototype heavy water reactor FUGEN is described. First, the outline of the core is introduced. The core management is generally the repetition of planning, practice and evaluation, but the evaluation is specifically important in FUGEN because FUGEN is a prototype reactor. In the reactor FUGEN, the fuel replacement plan which determines the number and position of fuels to be replaced, and fuel procurement plan based on the replacement plan are prepared. The control rod pattern is determined so that the thermal limit for the fuel assembly is secured throughout the fuel cycle, but the output flattening by control rods is scarcely necessary by adopting a distributed replacement method. After a replaced core has been composed, the maximum excess reactivity and reactivity shut-down margin are mainly measured at the start-up of the reactor to confirm the predetermined characteristics of the replaced core. The core life can be simply and accurately estimated by the measurement of /sup 10/B concentration in heavy water. The output distribution in the core is an important parameter for calculating the performance of the FUGEN reactor core. The output increasing procedure is also controlled in accordance with that of light water reactors.

  5. A contribution to the analysis of the thermal behaviour of Fast Breeder fuel rods with UO{sub 2}-PuO{sub 2} fuel; Contribucion al analisis del comportamiento termico de las barras combustibles de UO{sub 2}-PuO{sub 2} de los reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.; Elbel, H.

    1977-07-01

    The fuel of Fast Breeder Reactors which consists of Uranium and Plutonium dioxide is mainly characterized by the amount and distribution of void volume and Plutonium and the amount of oxygen. Irradiation experiments carried out with this fuel have shown that initial structure of the fuel pellet is subjected to large changes during operation. These are consequences of the radial and axial temperature gradients within the fuel rods. (Author) 54 refs.

  6. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  7. Design of virtual SCADA simulation system for pressurized water reactor

    Science.gov (United States)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  8. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  9. Novel Photocatalytic Reactor Development for Removal of Hydrocarbons from Water

    Directory of Open Access Journals (Sweden)

    Morgan Adams

    2008-01-01

    Full Text Available Hydrocarbons contamination of the marine environment generated by the offshore oil and gas industry is generated from a number of sources including oil contaminated drill cuttings and produced waters. The removal of hydrocarbons from both these sources is one of the most significant challenges facing this sector as it moves towards zero emissions. The application of a number of techniques which have been used to successfully destroy hydrocarbons in produced water and waste water effluents has previously been reported. This paper reports the application of semiconductor photocatalysis as a final polishing step for the removal of hydrocarbons from two waste effluent sources. Two reactor concepts were considered: a simple flat plate immobilised film unit, and a new rotating drum photocatalytic reactor. Both units proved to be effective in removing residual hydrocarbons from the effluent with the drum reactor reducing the hydrocarbon content by 90% under 10 minutes.

  10. Tridimensional ultrasonic images analysis for the in service inspection of fast breeder reactors; Analyse d'images tridimensionnelles ultrasonores pour l'inspection en service des reacteurs a neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Dancre, M

    1999-11-01

    Tridimensional image analysis provides a set of methods for the intelligent extraction of information in order to visualize, recognize or inspect objects in volumetric images. In this field of research, we are interested in algorithmic and methodological aspects to extract surface visual information embedded in volume ultrasonic images. The aim is to help a non-acoustician operator, possibly the system itself, to inspect surfaces of vessel and internals in Fast Breeder Reactors (FBR). Those surfaces are immersed in liquid metal, what justifies the ultrasonic technology choice. We expose firstly a state of the art on the visualization of volume ultrasonic images, the methods of noise analysis, the geometrical modelling for surface analysis and finally curves and surfaces matching. These four points are then inserted in a global analysis strategy that relies on an acoustical analysis (echoes recognition), an object analysis (object recognition and reconstruction) and a surface analysis (surface defects detection). Few literature can be found on ultrasonic echoes recognition through image analysis. We suggest an original method that can be generalized to all images with structured and non-structured noise. From a technical point of view, this methodology applied to echoes recognition turns out to be a cooperative approach between morphological mathematics and snakes (active contours). An entropy maximization technique is required for volumetric data binarization. (author)

  11. Chemical Gradients in Crud on Boiling Water Reactor Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Porter; D. E. Janney

    2007-04-01

    Crud (radioactive corrosion products formed inside nuclear reactors is a major problem in commercial power-producing nuclear reactors. Although there are numerous studies of simulated (non-radioactive) crud, characteristics of crud from actual reactors are rarely studied. This study reports scanning electron microscope (SEM) studies of fragments of crud from a commercially operating boiling water reactor. Chemical analyses in the SEM indicated that the crud closest to the outer surfaces of the fuel pins in some areas had Fe:Zn ratios close to 2:1, which decreased away from the fuel pin in some of the fragments. In combination with transmission electron microsope analyses (published elsewhere), these results suggest that the innermost layer of crud in some areas may consist of franklinite (ZnFe2O4, also called zinc spinel), while outer layers in these areas may be predominantly iron oxides.

  12. High-Temperature Water-Gas Shift Membrane Reactor Study

    Energy Technology Data Exchange (ETDEWEB)

    Ciocco, M.V.; Iyoha, O.; Enick, R.M.; Killmeyer, R.P.

    2007-06-01

    NETL’s Office of Research and Development is exploring the integration of membrane reactors into coal gasification plants as a way of increasing efficiency and reducing costs. Water-Gas Shift Reaction experiments were conducted in membrane reactors at conditions similar to those encountered at the outlet of a coal gasifier. The changes in reactant conversion and product selectivity due to the removal of hydrogen via the membrane reactor were quantified. Research was conducted to determine the influence of residence time and H2S on CO conversion in both Pd and Pd80wt%Cu membrane reactors. Effects of the hydrogen sulfide-to-hydrogen ratio on palladium and a palladium-copper alloy at high-temperature were also investigated. These results were compared to thermodynamic calculations for the stability of palladium sulfides.

  13. Key achievements in elementary R&D on water-cooled solid breeder blanket for ITER test blanket module in JAERI

    Science.gov (United States)

    Suzuki, S.; Enoeda, M.; Hatano, T.; Hirose, T.; Hayashi, K.; Tanigawa, H.; Ochiai, K.; Nishitani, T.; Tobita, K.; Akiba, M.

    2006-02-01

    This paper presents the significant progress made in the research and development (R&D) of key technologies on the water-cooled solid breeder blanket for the ITER test blanket modules in JAERI. Development of module fabrication technology, bonding technology of armours, measurement of thermo-mechanical properties of pebble beds, neutronics studies on a blanket module mockup and tritium release behaviour from a Li2TiO3 pebble bed under neutron-pulsed operation conditions are summarized. With the improvement of the heat treatment process for blanket module fabrication, a fine-grained microstructure of F82H can be obtained by homogenizing it at 1150 °C followed by normalizing it at 930 °C after the hot isostatic pressing process. Moreover, a promising bonding process for a tungsten armour and an F82H structural material was developed using a solid-state bonding method based on uniaxial hot compression without any artificial compliant layer. As a result of high heat flux tests of F82H first wall mockups, it has been confirmed that a fatigue lifetime correlation, which was developed for the ITER divertor, can be made applicable for the F82H first wall mockup. As for R&D on the breeder material, Li2TiO3, the effect of compression loads on effective thermal conductivity of pebble beds has been clarified for the Li2TiO3 pebble bed. The tritium breeding ratio of a simulated multi-layer blanket structure has successfully been measured using 14 MeV neutrons with an accuracy of 10%. The tritium release rate from the Li2TiO3 pebble has also been successfully measured with pulsed neutron irradiation, which simulates ITER operation.

  14. Utilization of plutonium in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    Japan's nuclear policy decides not to have excess plutonium. Upon assuming the future situation of the delay of FBR introduction, the JAERI performs the feasibility study of several types of the reduced-moderation water reactors (RMWRs). As the RMWRs have higher conversion ratio than LWRs, they are expected to enable multi-cycle utilization of plutonium, high burnup and long cycle operation, and enhancement of uranium resource utilization. While the full MOX LWRs are being developed, from viewpoint of suppressing the accumulation of plutonium, the RMWRs are thought to be more suitable. As plutonium inventory is larger in the RMWRs than in the full MOX LWRs, also from viewpoint of non-proliferation of nuclear materials, the RMWRs are thought to be more suitable. The current feasibility study will be performed until 2010 to confirm the position, to construct the reactor concept, and to demonstrate the feasibility on reactor physics and on thermal hydraulics. The present candidate reactor types of the study are three BWR types, heavy water cooled PWR type and light water cooled PWR type. Hereafter comprehensive evaluation from viewpoint of problems on fuel cycle, economy, continuity with conventional LWR technologies will be performed to extract the most suitable concept to satisfy the social needs and to construct the fundamental reactor concept to concentrate R and D effort. (K. Tsuchihashi)

  15. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  16. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  17. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    Science.gov (United States)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  18. Final Safety Analysis Addenda to Hazards Summary Report, Experimental Breeder Reactor II (EBR-II): upgrading of plant protection system. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    Allen, N. L.; Keeton, J. M.; Sackett, J. I. [comps.

    1980-06-01

    This report is the second in a series of compilations of the formal Final Safety Analysis Addenda (FSAA`s) to the EBR-II Hazard Summary Report and Addendum. Sections 2 and 3 are edited versions of the original FSAA`s prepared in support of certain modifications to the reactor-shutdown-system portion of the EBR-II plant-protection system. Section 4 is an edited version of the original FSAA prepared in support of certain modifications to a system classified as an engineered safety feature. These sections describe the pre- and postmodification system, the rationale for the modification, and required supporting safety analysis. Section 5 provides an updated description and analysis of the EBR-II emergency power system. Section 6 summarizes all significant modifications to the EBR-II plant-protection system to date.

  19. Final safety analysis addendum to hazard summary report, experimental breeder reactor No. II (EBR-II): the EBR-II cover-gas cleanup system

    Energy Technology Data Exchange (ETDEWEB)

    Fryer, R M; Monson, L R; Price, C C; Hooker, D W

    1979-04-01

    This report evaluates abnormal and accident conditions postulated for the EBR-II cover-gas cleanup system (CGCS). Major considerations include loss of CGCS function with a high level of cover-gas activity, loss of the liquid-nitrogen coolant required for removing fission products from the cover gas, contamination of the cover gas from sources other than the reactor, and loss of system pressure boundary. Calculated exposures resulting from the maximum hypothetical accident (MHA) are less than 2% of the 25-Rem limit stipulated in U.S. Regulation 10 CFR 100; i.e., a person standing at any point on an exclusion boundary (area radius of 600 m) for 2 h following onset of the postulated release would receive less than 0.45 Rem whole-body dose. The on-site whole-body dose (10 m from the source) would be less than 16 Rem.

  20. Microbial fouling community analysis of the cooling water system of a nuclear test reactor with emphasis on sulphate reducing bacteria.

    Science.gov (United States)

    Balamurugan, P; Joshi, M Hiren; Rao, T S

    2011-10-01

    Culture and molecular-based techniques were used to characterize bacterial diversity in the cooling water system of a fast breeder test reactor (FBTR). Techniques were selected for special emphasis on sulphate-reducing bacteria (SRB). Water samples from different locations of the FBTR cooling water system, in addition to biofilm scrapings from carbon steel coupons and a control SRB sample were characterized. Whole genome extraction of the water samples and SRB diversity by group specific primers were analysed using nested PCR and denaturing gradient gel electrophoresis (DGGE). The results of the bacterial assay in the cooling water showed that the total culturable bacteria (TCB) ranged from 10(3) to 10(5) cfu ml(-1); iron-reducing bacteria, 10(3) to 10(5) cfu ml(-1); iron oxidizing bacteria, 10(2) to 10(3) cfu ml(-1) and SRB, 2-29 cfu ml(-1). However, the counts of the various bacterial types in the biofilm sample were 2-3 orders of magnitude higher. SRB diversity by the nested PCR-DGGE approach showed the presence of groups 1, 5 and 6 in the FBTR cooling water system; however, groups 2, 3 and 4 were not detected. The study demonstrated that the PCR protocol influenced the results of the diversity analysis. The paper further discusses the microbiota of the cooling water system and its relevance in biofouling.

  1. Evaluation of fatigue data including reactor water environmental effects

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Nickell, R.E. [Applied Science and Technology, Poway, CA (United States); Van Der Sluys, W.A. [Alliance, OH (United States); Yukawa, S. [Boulder, CO (United States)

    2002-07-01

    Laboratory data have been gathered in the past decade indicating a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. The laboratory data under simulated operating conditions are being used to support arguments for revising the design-basis fatigue curves in the ASME Code Section III, Division 1, for Class 1 components. A thorough review of available laboratory fatigue data and their applicability to actual component operating conditions was performed. The evaluation divided the assembly, review and assessment of existing laboratory fatigue data and its applicability to plant operating conditions into four principal tasks: (1) review of available laboratory data relative to thresholds for environmental parameters, such as temperature, reactor water oxidation potential, strain rate, strain amplitude, reactor water flow rate, and component metal sulfur content; (2) determination of the relevance of the laboratory data to actual plant operating conditions; (3) review of laboratory S-N data curve-fitting models; and (4) assessment of existing ASME Code Section III Class 1 margins This paper summarizes the results of the data review. In addition, recommendations are made for additional laboratory testing intended to improve the applicability of laboratory test results under simulated reactor water environmental conditions. (authors)

  2. Materials Degradation in Light Water Reactors: Life After 60,???

    Energy Technology Data Exchange (ETDEWEB)

    Busby, Jeremy T [ORNL; Nanstad, Randy K [ORNL; Stoller, Roger E [ORNL; Feng, Zhili [ORNL; Naus, Dan J [ORNL

    2008-04-01

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase

  3. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  4. 78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

    Science.gov (United States)

    2013-09-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors... and operate integral pressurized water reactors (iPWR). This guidance applies to environmental reviews...

  5. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR) standard...

  6. Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bjerke, M.A.; Webster, C.C.

    1981-12-01

    Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations.

  7. Use of reactor effluent water as steam plant boiler feed

    Energy Technology Data Exchange (ETDEWEB)

    Clukey, H.V.

    1953-12-08

    The radiological aspects of a proposal to recover some of the heat now wasted in cooling water from the Hanford reactors by using the hot water as boiler feed for the steam plants in the 100 Areas are evaluated. The radioactive material in the hot effluent water will contaminate the boiler feed water system, cause additional radiation exposure of personnel, and increase the cost of maintenance and radiation protection, but very little radioactive material will be carried over into the steam system. At present steam loads, this proposal is economically attractive; other proposals being considered may nullify any savings from this one. 21 refs., 1 fig., 10 tabs.

  8. Computational Fluid Dynamics Analysis of Canadian Supercritical Water Reactor (SCWR)

    Science.gov (United States)

    Movassat, Mohammad; Bailey, Joanne; Yetisir, Metin

    2015-11-01

    A Computational Fluid Dynamics (CFD) simulation was performed on the proposed design for the Canadian SuperCritical Water Reactor (SCWR). The proposed Canadian SCWR is a 1200 MW(e) supercritical light-water cooled nuclear reactor with pressurized fuel channels. The reactor concept uses an inlet plenum that all fuel channels are attached to and an outlet header nested inside the inlet plenum. The coolant enters the inlet plenum at 350 C and exits the outlet header at 625 C. The operating pressure is approximately 26 MPa. The high pressure and high temperature outlet conditions result in a higher electric conversion efficiency as compared to existing light water reactors. In this work, CFD simulations were performed to model fluid flow and heat transfer in the inlet plenum, outlet header, and various parts of the fuel assembly. The ANSYS Fluent solver was used for simulations. Results showed that mass flow rate distribution in fuel channels varies radially and the inner channels achieve higher outlet temperatures. At the outlet header, zones with rotational flow were formed as the fluid from 336 fuel channels merged. Results also suggested that insulation of the outlet header should be considered to reduce the thermal stresses caused by the large temperature gradients.

  9. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  10. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  11. Upper internals arrangement for a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  12. COMSORS: A light water reactor chemical core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States). Chemical Technology Div.; Kenton, M.A. [Creare Inc., Hanover, NH (United States)

    1997-02-24

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B{sub 2}O{sub 3}) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation.

  13. 用于池式快堆系统分析的钠池三维模型开发%Development of Three-Dimensional Sodium Pool Model for System Analysis of Pool-Type Liquid Metal Fast Breeder Reactor

    Institute of Scientific and Technical Information of China (English)

    隋丹婷; 陆道纲; 张盼

    2012-01-01

    由于池式快堆钠池内的热工水力学特性对反应堆的安全运行有重要影响,本文采用基于交错网格的SIMPLE算法开发直角坐标系和柱坐标系下钠池三维计算软件.应用CFX软件进行验证之后,完成了三维流场分析程序与系统分析软件SAC-CFR的耦合,并用耦合后的程序分析日本文殊快堆45%功率稳态运行工况上腔室内的流场分布,初步验证了堆芯上腔三维化的SAC-CFR用于系统分析的有效性,为进一步开发事故模型、非能动余热排出系统模型做准备.%As the thermal-hydraulic characteristic in sodium pool is crucial for safety operation of liquid metal fast breeder reactor (LMFBR), a three-dimensional sodium pool thermal-hydraulic analysis code was developed based on SIMPLE algorithm on stagger grid under Cartesian coordinates and cylindrical coordinates. After the validation with CFX, coupling between the analysis code and SAC-CFR was completed) and then the coupled code was applied to the flow field analysis in upper plenum of Monju Plant at 45% thermal power steady-state operation condition, which preliminary shows the effectiveness of the system analysis with coupled code and makes preparations for further development of accident analysis model and passive residual heat removal system.

  14. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  15. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design.... Advanced Boiling Water Reactor (U.S. ABWR) standard plant design to comply with the NRC's aircraft impact...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis A...

  16. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security Requirements 1.0 Background The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by the..., which utilized a forced-circulation, direct-cycle boiling water reactor as its heat source. The plant is...

  17. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis A...

  18. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  19. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  20. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  1. Optofluidic planar reactors for photocatalytic water treatment using solar energy

    Science.gov (United States)

    Lei, Lei; Wang, Ning; Zhang, X. M.; Tai, Qidong; Tsai, Din Ping; Chan, Helen L. W.

    2010-01-01

    Optofluidics may hold the key to greater success of photocatalytic water treatment. This is evidenced by our findings in this paper that the planar microfluidic reactor can overcome the limitations of mass transfer and photon transfer in the previous photocatalytic reactors and improve the photoreaction efficiency by more than 100 times. The microreactor has a planar chamber (5 cm×1.8 cm×100 μm) enclosed by two TiO2-coated glass slides as the top cover and bottom substrate and a microstructured UV-cured NOA81 layer as the sealant and flow input∕output. In experiment, the microreactor achieves 30% degradation of 3 ml 3×10−5M methylene blue within 5 min and shows a reaction rate constant two orders higher than the bulk reactor. Under optimized conditions, a reaction rate of 8% s−1 is achieved under solar irradiation. The average apparent quantum efficiency is found to be only 0.25%, but the effective apparent quantum efficiency reaches as high as 25%. Optofluidic reactors inherit the merits of microfluidics, such as large surface∕volume ratio, easy flow control, and rapid fabrication and offer a promising prospect for large-volume photocatalytic water treatment. PMID:21267436

  2. Multi-Applications Small Light Water Reactor - NERI Final Report

    Energy Technology Data Exchange (ETDEWEB)

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  3. Gravity Scaling of a Power Reactor Water Shield

    Science.gov (United States)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  4. Silicon carbide composite for light water reactor fuel assembly applications

    Science.gov (United States)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  5. Silicon carbide composite for light water reactor fuel assembly applications

    Energy Technology Data Exchange (ETDEWEB)

    Yueh, Ken, E-mail: kyueh@epri.com [Fuel Reliability Program, EPRI, 1300 West WT Harris Blvd, Charlotte, NC 28262 (United States); Terrani, Kurt A., E-mail: terranika@ornl.gov [Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Rd. MS 6093, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The feasibility of using SiC{sub f}–SiC{sub m} composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  6. Component test for validation of the design life of reactor vessel wall of the fast breeder reactor SNR 300 regarding creep-fatigue. Modellversuch zur Absicherung der Auslegung der Reaktortankwand des SNR 300 hinsichtlich Kriechermuedung

    Energy Technology Data Exchange (ETDEWEB)

    Lohse, U. (Interatom GmbH, Bergisch Gladbach (Germany)); Laue, H. (Interatom GmbH, Bergisch Gladbach (Germany)); Rathjen, P. (Interatom GmbH, Bergisch Gladbach (Germany)); Maile, K. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany)); Eckert, W. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany)); Purper, H. (Staatliche Materialpruefungsanstalt, Stuttgart (Germany))

    1991-01-01

    The design of the reactor vessel wall is tested under a long-term creep-fatigue stress, with the aid of a similar components test. For this purpose, the results of calculation are compared with the experimental results concerning deformation depending on point and time, and damage depending on the initial state of material, point and time. (orig./HP)

  7. Risks of nuclear energy technology safety concepts of light water reactors

    CERN Document Server

    Kessler, Günter; Schlüter, Franz-Hermann

    2014-01-01

    The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on?reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: ? A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Ch

  8. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  9. Spiral-shaped reactor for water disinfection

    KAUST Repository

    Soukane, Sofiane

    2016-04-20

    Chlorine-based processes are still widely used for water disinfection. The disinfection process for municipal water consumption is usually carried out in large tanks, specifically designed to verify several hydraulic and disinfection criteria. The hydrodynamic behavior of contact tanks of different shapes, each with an approximate total volume of 50,000 m3, was analyzed by solving turbulent momentum transport equations with a computational fluid dynamics code, namely ANSYS fluent. Numerical experiments of a tracer pulse were performed for each design to generate flow through curves and investigate species residence time distribution for different inlet flow rates, ranging from 3 to 12 m3 s−1. A new nature-inspired Conch tank design whose shape follows an Archimedean spiral was then developed. The spiral design is shown to strongly outperform the other tanks’ designs for all the selected plug flow criteria with an enhancement in efficiency, less short circuiting, and an order of magnitude improvement in mixing and dispersion. Moreover, following the intensification philosophy, after 50% reduction in its size, the new design retains its properties and still gives far better results than the classical shapes.

  10. Dense Medium Plasma Water Purification Reactor (DMP WaPR) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Dense Medium Plasma Water Purification Reactor offers significant improvements over existing water purification technologies used in Advanced Life Support...

  11. Wear resistant zirconium base alloy article for water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gillett, J.E.; Shockling, L.A.; Sherwood, D.G.

    1988-03-01

    In a water reactor operating environment, the combination having improved fretting wear resistance is described comprising: an elongated tubular water displacer rod; having a low neutron absorption cross section guide support plates distributed along the length of the water displacer rod; the water displacer rod intersecting the guide support plates through apertures in the guide support plates; the water displacer rod having a plurality of spaced apart annular electrospark deposited coatings, each coating facing the wall of a respective aperture, the electrospark deposited coatings comprising Cr/sub 2/C/sub 3/; wherein the water displacer rod has a tube wall composed of a zirconium base alloy; and wherein the guide support plates are composed of a stainless steel alloy.

  12. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  13. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  14. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  15. Water-moderated reactor fuel cladding reliability study

    OpenAIRE

    Бакутяк, Елена Викторовна; Пелых, Сергей Николаевич

    2014-01-01

    Considering the fuel element, averaged by fuel assembly (FA) of water-moderated reactor with the power of 1000 MW (VVER-1000), the number of fuel elements with the greatest cladding failure probability after 4 operation years at Khmelnitsky NPP-2 (KNPP-2) is found. This will allow to calculate the fuel cladding failure probability and determine the most likely cladding damages, which will enable to improve the performance and economic indexes of VVER.The novelty of the paper lies in calculati...

  16. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems

    NARCIS (Netherlands)

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.

    In

  17. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems.

    NARCIS (Netherlands)

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.In this the

  18. Breeder Spent Fuel Handling Program multipurpose cask design basis document

    Energy Technology Data Exchange (ETDEWEB)

    Duckett, A.J.; Sorenson, K.B.

    1985-09-01

    The Breeder Spent Fuel Handling (BSFH) Program multipurpose cask Design Basis Document defines the performance requirements essential to the development of a legal weight truck cask to transport FFTF spent fuel from reactor to a reprocessing facility and the resultant High Level Waste (HLW) to a repository. 1 ref.

  19. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor.

  20. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  1. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR...

  2. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems

    OpenAIRE

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.

    In this thesis the possibilities are presented for fixed-film post-treatment of anaerobically digested domestic sewage and water reconditioning in aquacultural water recirculation systems. Emphasis i...

  3. Simulation of Water Gas Shift Zeolite Membrane Reactor

    Science.gov (United States)

    Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

    2017-07-01

    The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

  4. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  5. Non-linear analysis in Light Water Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  6. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B. [University of Toronto, Centre for Nuclear Engineering, Toronto, Ontario (Canada)

    1997-07-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  7. Transactions of the nineteenth water reactor safety information meeting

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. (comp.)

    1991-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

  8. Materials Inventory Database for the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  9. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  10. Numerical study of the effects of surface roughness on water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu; Ahmad, Sarfraz; Cho, Jinsoo

    2016-04-01

    UV reactors are an emerging choice as a big barrier against the pathogens present in drinking water. However, the precise role of reactor's wall roughness for cross flow ultraviolet (CF-UV) and axial flow ultraviolet (AF-UV) water disinfection reactors are unknown. In this paper, the influences of reactor's wall roughness were investigated with a view to identify their role on the performance factors namely dose distribution and reduction equivalent dose (RED). Herein, the relative effects of reactor's wall roughness on the performance of CF-UV and AF-UV reactors were also highlighted. This numerical study is a first step towards the comprehensive analysis of the effects of reactor's wall roughness for UV reactor. A numerical analysis was performed using ANSYS Fluent 15 academic version. The reactor's wall roughness has a significant effect on the RED. We found that the increase in RED is Reynolds number dependent (at lower value of turbulent Reynolds number the effects are remarkable). The effects of reactor's roughness were more pronounced for AF-UV reactor. The simulation results suggest that the study of reactor's wall roughness provides valuable insight to fully understand the effects of reactor's wall roughness and its impact on the flow behavior and other features of CF-UV and AF-UV water disinfection reactors. Copyright © 2016 Elsevier Ltd. All rights reserved.

  11. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  12. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  13. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    1980-11-01

    This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  14. The use of PRA in the development of ALWR (advanced light water reactor) design requirements. [Advanced Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Summitt, R.L. (Safety and Reliability Optimization Services, Inc., Knoxville, TN (USA)); Additon, S.L. (TENERA, L.P., Bethesda, MD (USA)); Pasedag, W.F. (USDOE, Washington, DC (USA))

    1989-01-01

    The current hiatus in nuclear power plant orders provides an opportunity for the development of advanced light water reactor (ALWR) design concepts and regulatory requirements which incorporate the insights gained from the application of the probabilistic risk assessment. The US Department of Energy is assisting the Electric Power Research Institute (EPRI) in the incorporation of PRA insights into the specification of the utility requirements, and reactor vendors in support of the conceptual design of safety systems, for such advanced plants. This paper reviews the applications of PRA methods in this development of specifications for, and the design of simplified, rugged ALWRs with a significantly improved risk profile. Specific examples of the impact of utilizing published PRA insights, construction and use of functional PRA models, and feedback of PRA experience into the specification of the key assumptions and groundrules for ALWR PRAs are presented. 13 refs., 3 tabs.

  15. Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

  16. Camera Inspection Arm for Boiling Water Reactors - 13330

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Scott; Rood, Marc [S.A. Technology, 3985 S. Lincoln Ave, Loveland, CO 80537 (United States)

    2013-07-01

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  17. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  18. Supercritical Water Reactor Cycle for Medium Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  19. Characterization of 14C in Swedish light water reactors.

    Science.gov (United States)

    Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

    2008-08-01

    This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

  20. Boiling-Water Reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

  1. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  2. Structural Integrity of Water Reactor Pressure Boundary Components.

    Science.gov (United States)

    1980-08-01

    tests of reference steels of the NRC light water reactor, pressure vessel irradiation dosimetry program. SECURITY CLAS5IICATION 0PHiS PA6GMbn" Dfat ...multiple specimen R- curve approach; NRL emphasis was on the SSC procedure as it is being developed for hot- cell testing of irradiated materials. MULTIPLE...a second autoclave, capable of testing 50 or 100 mm (2T or 4T) thick CT or WOL specimens, was installed in a hot cell and a test was started on 2T-CT

  3. 77 FR 16098 - In the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II...

    Science.gov (United States)

    2012-03-19

    ... the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II Containments... operate boiling-water reactors (BWRs) with Mark I and Mark II containment designs. II On March 11, 2011, a... Nuclear Reactor Regulation. Operating Boiling Water Reactor Licenses With Mark I and Mark II Containments...

  4. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  5. Continuous-flow solar UVB disinfection reactor for drinking water.

    Science.gov (United States)

    Mbonimpa, Eric Gentil; Vadheim, Bryan; Blatchley, Ernest R

    2012-05-01

    Access to safe, reliable sources of drinking water is a long-standing problem among people in developing countries. Sustainable solutions to these problems often involve point-of-use or community-scale water treatment systems that rely on locally-available resources and expertise. This philosophy was used in the development of a continuous-flow, solar UVB disinfection system. Numerical modeling of solar UVB spectral irradiance was used to define temporal variations in spectral irradiance at several geographically-distinct locations. The results of these simulations indicated that a solar UVB system would benefit from incorporation of a device to amplify ambient UVB fluence rate. A compound parabolic collector (CPC) was selected for this purpose. Design of the CPC was based on numerical simulations that accounted for the shape of the collector and reflectance. Based on these simulations, a prototype CPC was constructed using materials that would be available and inexpensive in many developing countries. A UVB-transparent pipe was positioned in the focal area of the CPC; water was pumped through the pipe to allow exposure of waterborne microbes to germicidal solar UVB radiation. The system was demonstrated to be effective for inactivation of Escherichia coli, and DNA-weighted UV dose was shown to govern reactor performance. The design of the reactor is expected to scale linearly, and improvements in process performance (relative to results from the prototype) can be expected by use of larger CPC geometry, inclusion of better reflective materials, and application in areas with greater ambient solar UV spectral irradiance than the location of the prototype tests. The system is expected to have application for water treatment among communities in (developing) countries in near-equatorial and tropical locations. It may also have application for disaster relief or military field operations, as well as in water treatment in areas of developed countries that receive

  6. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Paul Y [Los Alamos National Laboratory

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  7. Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

    OpenAIRE

    2009-01-01

    A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (S...

  8. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  9. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  10. Reactor vessel water level estimation during severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-06-15

    Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

  11. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  12. Tritium distribution modeling in a Light Water New Production Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jaeckle, J.W.

    1989-05-01

    The tritium distribution and tritium release pathways in a new light water production reactor were examined. A computer model was developed to track the tritium as it makes its way through the various plant systems and ends up either as a release to the atmosphere, the cooling tower blowdown or to the solid waste system. The model was designed to predict the integrated yearly tritium releases and provide estimated airborne tritium concentrations in various locations within the plant. WNP-1 was used as a representative model for a Light Water New Production Reactor (LWNPR). The Tritium Distribution Model solves for the time dependent tritium concentration in a system of nodes. These nodes are connected to one another via a set of internodal flow paths and to various sources and sinks. For example, plant systems such as the primary system are the nodes, piping and leaks are the internodal flow paths, make-up water is a source, and release to the atmosphere is a sink. The expected water mass of each node; the flow rates between nodes, sources, and sinks; and tritium source rates are provided as input. The code will solve for the time dependent tritium concentration in each node and the amount of tritium ''released'' to the sinks. Preliminary calculations have been performed using WNP-1 plant specific information obtained primarily from the WNP-1 FSAR. Further work is currently in progress to refine the model and provide a more realistic set of input values which will better represent an operating LWNPR. 1 ref., 1 fig., 1 tab.

  13. Continuous supercritical water gasification of isooctane: A promising reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Susanti, Ratna F.; Kim, Jae-Duck; Kim, Jaehoon [Supercritical Fluid Research Laboratory, Clean Energy Center, Energy Division, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seoungbuk-gu, Seoul 136-791 (Korea); Department of Green Process and System Engineering, University of Science and Technology (UST), 113 Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea); Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Energy Center, Energy Division, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seoungbuk-gu, Seoul 136-791 (Korea); Lee, Youn-Woo [School of Chemical and Biological Engineering, Seoul National University, Gwanangro 599, Gwanak-gu, Seoul 151-744 (Korea)

    2010-03-15

    A new design of supercritical water gasification system was developed to achieve high hydrogen gas yield and good gas-liquid flow stability. The apparatus consisted of a reaction zone, an insulation zone and a cooling zone that were directly connected to the reaction zone. The reactor was set up at an inclination of 75 from vertical position, and feed and water were introduced at the bottom of the reactor. The performances of this new system were investigated with gasification of isooctane at various experimental conditions - reaction temperatures of 601-676 C, residence times of 6-33 s, isooctane concentrations of 5-33 wt%, and oxidant (hydrogen peroxide) concentrations up to 4507 mmol/L without using catalysts. A significant increase in hydrogen gas yield, almost four times higher than that from the previous up-down gasifier configuration (B. Veriansyah, J. Kim, J.D. Kim, Y.W. Lee, Hydrogen Production by Gasification of Isooctane using Supercritical Water, Int. J. Green Energy. 5 (2008) 322-333) was observed with the present gasifier configuration. High hydrogen gas yield (6.13 mol/mol isooctane) was obtained at high reaction temperature of 637 C, a low feed concentration of 9.9 wt% and a long residence time of 18 s in the presence of 2701.1 mmol/L hydrogen peroxide. At this condition, the produced gases mainly consisted of hydrogen (59.5 mol%), methane (14.8 mol%) and carbon dioxide (22.0 mol%), and a small amount of carbon monoxide (1.6 mol%) and C{sub 2}-C{sub 3} species (2.1 mol%). Reaction mechanisms of supercritical water gasification of isooctane were also presented. (author)

  14. Fast-Mixed Spectrum Reactor. Progress report for 1979

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.J.; Cerbone, R.J.

    1980-05-01

    This report summarizes the progress of the Fast Mixed Spectrum Reactor (FMSR) since the publication of the Interim Report in January 1979. The FMSR program was initiated to determine the feasibility of a breeder reactor concept which operated on a once-through-and-store fuel cycle and for which the only feed would be natural uranium. A first or startup core enriched to a maximum of about eleven percent in uranium-235 would be required. The concept has excellent antiproliferation advantages. In the once-through and store mode, the FMSR has a resource utilization which is a factor of four higher than a light water reactor.

  15. Revised accident source terms for light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  17. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  18. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  19. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&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.

  20. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

    2015-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&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.

  1. Pressurized-water reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  2. Technologies for Upgrading Light Water Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  3. Aging study of boiling water reactor high pressure injection systems

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  4. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  5. A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

    2002-03-15

    Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

  6. Development of Novel Water-Gas Shift Membrane Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ho, W. S. Winston

    2004-12-29

    This report summarizes the objectives, technical barrier, approach, and accomplishments for the development of a novel water-gas-shift (WGS) membrane reactor for hydrogen enhancement and CO reduction. We have synthesized novel CO{sub 2}-selective membranes with high CO{sub 2} permeabilities and high CO{sub 2}/H{sub 2} and CO{sub 2}/CO selectivities by incorporating amino groups in polymer networks. We have also developed a one-dimensional non-isothermal model for the countercurrent WGS membrane reactor. The modeling results have shown that H{sub 2} enhancement (>99.6% H{sub 2} for the steam reforming of methane and >54% H{sub 2} for the autothermal reforming of gasoline with air on a dry basis) via CO{sub 2} removal and CO reduction to 10 ppm or lower are achievable for synthesis gases. With this model, we have elucidated the effects of system parameters, including CO{sub 2}/H{sub 2} selectivity, CO{sub 2} permeability, sweep/feed flow rate ratio, feed temperature, sweep temperature, feed pressure, catalyst activity, and feed CO concentration, on the membrane reactor performance. Based on the modeling study using the membrane data obtained, we showed the feasibility of achieving H{sub 2} enhancement via CO{sub 2} removal, CO reduction to {le} 10 ppm, and high H{sub 2} recovery. Using the membrane synthesized, we have obtained <10 ppm CO in the H{sub 2} product in WGS membrane reactor experiments. From the experiments, we verified the model developed. In addition, we removed CO{sub 2} from a syngas containing 17% CO{sub 2} to about 30 ppm. The CO{sub 2} removal data agreed well with the model developed. The syngas with about 0.1% CO{sub 2} and 1% CO was processed to convert the carbon oxides to methane via methanation to obtain <5 ppm CO in the H{sub 2} product.

  7. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  8. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  9. Aging assessment of PWR (Pressurized Water Reactor) Auxiliary Feedwater Systems

    Energy Technology Data Exchange (ETDEWEB)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab.

  10. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) requesting exemptions from certain security requirements in Title 10 of the Code Federal Regulations (10 CFR) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has...

  11. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor...

  12. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  13. Final Report on Isotope Ratio Techniques for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

    2009-07-01

    The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

  14. Application of fully ceramic microencapsulated fuels in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, C.; George, N.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States); Godfrey, A.; Terrani, K.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01

    This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

  15. Application of Fully Ceramic Microencapsulated Fuels in Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL

    2012-01-01

    This study aims to perform a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in Light Water Reactors (LWRs). In particular pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor. Using uranium-based fuel and transuranic (TRU) based fuel in TRistructural ISOtropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher physical density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the TRU based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, feasibility of core designs fully loaded with TRU FCM lattices was demonstrated using the NESTLE three-dimensional core simulator.

  16. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  17. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    Energy Technology Data Exchange (ETDEWEB)

    None

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  18. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  19. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  20. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-05-01

    proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

  1. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  2. Nuclear reactor for breeding U.sup.233

    Science.gov (United States)

    Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding U.sup.233 for use in a light-water breeder reactor includes physically separated regions containing U.sup.235 fissile material and U.sup.238 fertile material and Th.sup.232 fertile material and Pu.sup.239 fissile material, if available. Preferably the U.sup.235 fissile material and U.sup.238 fertile material are contained in longitudinally movable seed regions and the Pu.sup.239 fissile material and Th.sup.232 fertile material are contained in blanket regions surrounding the seed regions.

  3. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  4. Synergistic Water-Treatment Reactors Using a TiO2-Modified Ti-Mesh Filter

    Directory of Open Access Journals (Sweden)

    Akira Fujishima

    2013-07-01

    Full Text Available The recent applications of a TiO2-modified Ti-mesh filter (TMiP™ for water purification are summarized with newly collected data including biological assays as well as sewage water treatment. The water purification reactors consist of the combination of a TMiP, a UV lamp, an excimer VUV lamp, and an ozonation unit. The water purification abilities of the reactor were evaluated by decomposition of organic contaminants, inactivation of waterborne pathogens, and treatment efficiency for sewage water. The UV-C/TMiP/O3 reactor disinfected E. coli in aqueous suspension in approximately 1 min completely, and also decreased the number of E. coli in sewage water in 15 min dramatically. The observed rate constants of 7.5 L/min and 1.3 L/min were calculated by pseudo-first-order kinetic analysis respectively. Although organic substances in sewage water were supposed to prevent the UV-C/TMiP/O3 reactor from purifying water, the reactor reduced E. coli in sewage water continuously. On the other hand, although much higher efficiencies for decomposition of organic pollutants in water were achieved in the excimer/TMiP reactor, the disinfection activity of the reactor for waterborne pathogens was not as effective as the other reactors. The difference of efficiency between organic pollutants and waterborne pathogens in the excimer/TMiP reactor may be due to the size, the structure, and the decomposition mechanism of the organic pollutants and waterborne pathogens. These results show that a suitable system assisted by synergy of photocatalysts and other technologies such as ozonation has a huge potential as a practical wastewater purification system.

  5. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  6. Grey water treatment in upflow anaerobic sludge blanket (UASB) reactor at different temperatures.

    Science.gov (United States)

    Elmitwalli, Tarek; Otterpohl, Ralf

    2011-01-01

    The treatment of grey water in two upflow anaerobic sludge blanket (UASB) reactors, operated at different hydraulic retention times (HRTs) and temperatures, was investigated. The first reactor (UASB-A) was operated at ambient temperature (14-25 degrees C) and HRT of 20, 12 and 8 h, while the second reactor (UASB-30) was operated at controlled temperature of 30 degrees C and HRT of 16, 10 and 6 h. The two reactors were fed with grey water from 'Flintenbreite' settlement in Luebeck, Germany. When the grey water was treated in the UASB reactor at 30 degrees C, total chemical oxygen demand (CODt) removal of 52-64% was achieved at HRT between 6 and 16 h, while at lower temperature lower removal (31-41%) was obtained at HRT between 8 and 20 h. Total nitrogen and phosphorous removal in the UASB reactors were limited (22-36 and 10-24%, respectively) at all operational conditions. The results showed that at increasing temperature or decreasing HRT of the reactors, maximum specific methanogenic activity of the sludge in the reactors improved. As the UASB reactor showed a significantly higher COD removal (31-64%) than the septic tank (11-14%) even at low temperature, it is recommended to use UASB reactor instead of septic tank (the most common system) for grey water pre-treatment. Based on the achieved results and due to high peak flow factor, a HRT between 8 and 12 h can be considered the suitable HRT for the UASB reactor treating grey water at temperature 20-30 degrees C, while a HRT of 12-24 h can be applied at temperature lower than 20 degrees C.

  7. Multi-Application Small Light Water Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept

  8. Multi-Application Small Light Water Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept

  9. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  10. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  11. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  12. Bacterial Colonization of Pellet Softening Reactors Used during Drinking Water Treatment

    NARCIS (Netherlands)

    Hammes, F.; Boon, N.; Vital, M.; Ross, P.; Magic-Knezev, A.; Dignum, M.

    2010-01-01

    Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase t

  13. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the ADAMS... published a notice requesting public comments on draft LR-ISG-2011-04, ``Updated Aging Management...

  14. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  15. Impact of inflow transport approximation on light water reactor analysis

    Science.gov (United States)

    Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung

    2015-10-01

    The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.

  16. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors; Caracterisation microstructurale et modelisation du durcissement des aciers austenitiques irradies des structures internes des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C

    2003-07-01

    The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

  17. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  18. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  19. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  20. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  1. Experimental and analytical study on thermal hydraulics in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Araya, Fumimasa; Ohnuki, Akira; Yoshida, Hiroyuki; Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Study and development of reduced-moderation spectrum water reactor proceeds as a option of the future type reactor in Japan Atomic Energy Research Institute (JAERI). The reduced-moderation spectrum in which a neutron has higher energy than the conventional water reactors is achieved by decreasing moderator-to-fuel ratio in the lattice core of the reactor. Conversion ratio in the reduced-moderation water reactor can be more than 1.0. High burnup and long term cycle operation of the reactor are expected. A type of heavy water cooled PWR and three types of BWR are discussed as follows; For the PWR, (1) critical heat flux experiments in hexagonal tight lattice core, (2) evaluation of cooling limit at a nominal power operation, and (3) analysis of rewetting cooling behavior at loss of coolant accident following with large scale pipe rupture. For the BWR, analyses of cooling limit at a nominal power operation of, (1) no blanket BWR, (2) long term cycle operation BWR, and (3) high conversion ratio BWR. The experiments and the analyses proved that the basic thermal hydraulic characteristics of these reduced-moderation water reactors satisfy the essential points of the safety requirements. (Suetake, M.)

  2. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    Science.gov (United States)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  3. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  4. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  5. Pre-conceptual design study on K-DEMO ceramic breeder blanket

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Sung, E-mail: jspark@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kwon, Sungjin; Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Brown, Thomas; Neilson, George [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2015-11-15

    A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (B{sub T0} = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li{sub 4}SiO{sub 4} pebbles with Be{sub 12}Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials.

  6. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  7. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  8. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    Science.gov (United States)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  9. Passive gamma analysis of the boiling-water-reactor assemblies

    Science.gov (United States)

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; Vaccaro, S.

    2016-09-01

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative-Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  10. Passive gamma analysis of the boiling-water-reactor assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  11. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  12. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  13. Performance of materials in the component cooling water systems of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.

    1993-06-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed.

  14. Design and installation of a hot water layer system at the Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila

    2013-01-01

    Full Text Available A hot water layer system (HWLS is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

  15. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  16. Modeling of Fischer-Tropsch Synthesis in a Slurry Reactor with Water Permeable Membrane

    Institute of Scientific and Technical Information of China (English)

    Fabiano A. N. Fernandes

    2007-01-01

    Fischer-Tropsch synthesis is an important chemical process for the production of liquid fuels and olefins. In recent years, the abundant availability of natural gas and the increasing demand of olefins, diesel, and waxes have led to a high interest to further develop this process. A mathematical model of a slurry membrane reactor used for syngas polymerization was developed to simulate and compare the maximum yields and operating conditions in the reactor with that in a conventional slurry reactor.The carbon polymerization was studied from a modeling point of view in a slurry reactor with a water permeable membrane and a conventional slurry reactor. Simulation results show that different parameters affect syngas conversion and carbon product distribution, such as the hydrogen to carbon monoxide ratio,and the membrane parameters such as membrane permeance.

  17. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  18. Waste management of tar water from pyrolysis and gasification of biomass in biogas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mogensen, A.S.; Schmidt, J.E.; Angelidaki, R.; Ahring, B.K.

    1998-08-01

    The digestion and detoxification of pyrolysis condensate and wet oxidised pyrolysis condensate was studied in different reactor systems: combined anaerobic and denitrifying UASB reactors, conventional UASB reactors and CSTR`s. The pyrolysis condensate and the wet oxidised condensate have a biogas potential of 190 m{sup 3}/ton VS, and the low amount of suspended solids is allowing the waste water to be treated in the UASB reactor as well as in the CSTR. The pyrolysis condensate could successfully be degraded in a CSTR in a 5% concentration when co-digested with manure, and the wet oxidised pyrolysis condensate could be degraded when added at a concentration of 30%. The UASB reactor was preferred over the CSTR since the xenobiotic compounds present in the waste water might easily be absorbed in the co-substrate required when using the CSTR technology. Consequently, decreased degradation of xenobiotics would be observed in the CSTR. A combined anaerobic and denitrifying UASB reactor was successfully digesting 5.5% of wet oxidised pyrolysis condensate, but further loading increments deteriorated the anaerobic digestion process. However, when a UASB reactor was fed with pyrolysis condensate (up to 100%) good reactor operation was observed indicating that the waste could be used as substrate in the biogas process, even in very high concentrations. The detoxification of pyrolysis condensate was further studied and the toxicity of pyrolysis condensate was decreased more than 77 times in the UASB reactor that was operating on 100% pyrolysis condensate. Phenol, methyl and dimethyl phenols along with methoxyphenols were shown to be degraded within the rector systems. Degradation rates for phenol and substituted phenols were determined indicating that the biomass was selective towards the substrates. Maximum growth rates and half saturation constants for phenol, 4-Methylphenol and 2-Methoxy-4-methylphenol were determined in batch experiments. A UASB reactor concept was further

  19. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  20. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  1. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  2. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    Energy Technology Data Exchange (ETDEWEB)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    2001-07-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 {approx} 10{sup -V} at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  3. Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1984-07-01

    It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

  4. Solar disinfection of contaminated water: a comparison of three small-scale reactors

    Energy Technology Data Exchange (ETDEWEB)

    McLoughlin, O.A.; Gill, L.W. [Dublin Univ. (Ireland). Dept. of Civil, Structural and Environmental Engineering; Kehoe, S.C. [Royal College of Surgeons in Ireland, Dublin (Ireland). Dept. of Surgery; McGuigan, K.G.; Duffy, E.F.; Al Touati, F. [Royal College of Surgeons in Ireland, Dublin (Ireland). Dept. of Physiology and Medical Physics; Gernjak, W.; Alberola, I.O.; Malato Rodriguez, S. [Plataforma Solar de Almeria (CIEMAT), Tabemas (Spain)

    2004-11-01

    This paper compares three different collector shapes for the disinfection of water heavily contaminated with Escherichia coli (K-12). Tests were carried out in real sunlight using laboratory scale reactors to determine the performance of different reflector profiles. The reactors were constructed using Pyrex tubing and aluminium reflectors of compound parabolic, parabolic and V-groove profiles. Results have shown that the compound parabolic reflector promoted a more successful inactivation of E. coli than the parabolic and V-groove profiles. Tests were also carried out to assess the improvement to disinfection which could be achieved using TiO{sub 2} coated Pyrex rods fixed within the reactors. This technique, however, yielded a slight enhancement in the compound parabolic reactor but no benefit to overall disinfection performance in either the parabolic or V-groove reactors. These results show that the use of UV sunlight to disinfect contaminated drinking water in a full-scale continuous flow solar reactor is both promising and an appropriate technology for developing countries but that the inclusion of a fixed photocatalyst within the reactor tubes has yet to prove any significant improvement. (Author)

  5. Searching for full power control rod patterns in a boiling water reactor using genetic algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jlmt@nuclear.inin.mx; Ortiz, Juan Jose [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: jjortiz@nuclear.inin.mx; Requena, Ignacio [Departamento Ciencias Computacion e I.A. ETSII, Informatica, Universidad de Granada, C. Daniel Saucedo Aranda s/n. 18071 Granada (Spain)]. E-mail: requena@decsai.ugr.es; Perusquia, Raul [Departamento Sistemas Nucleares, ININ, Carr. Mexico-Toluca Km. 36.5, Ocoyoacac, Edo. de Mexico (Mexico)]. E-mail: rpc@nuclear.inin.mx

    2004-11-01

    One of the most important questions related to both safety and economic aspects in a nuclear power reactor operation, is without any doubt its reactivity control. During normal operation of a boiling water reactor, the reactivity control of its core is strongly determined by control rods patterns efficiency. In this paper, GACRP system is proposed based on the concepts of genetic algorithms for full power control rod patterns search. This system was carried out using LVNPP transition cycle characteristics, being applied too to an equilibrium cycle. Several operation scenarios, including core water flow variation throughout the cycle and different target axial power distributions, are considered. Genetic algorithm fitness function includes reactor security parameters, such as MLHGR, MCPR, reactor k{sub eff} and axial power density.

  6. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  7. Low-temperature water reactor for the district heating atomic power plant

    Energy Technology Data Exchange (ETDEWEB)

    Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Y.V.

    1978-04-01

    A natural convection low-pressure water reactor can be utilized as a source of district heating. This provides inherent safety factors under conditions requiring emergency core cooling. The reactor pressure vessel is contained within a prestressed concrete shell, both of which are designed to withstand accident overpressure. This also results in a relatively thin-walled reactor vessel that can be fabricated on-site. The overall safety and economy of such a system merits further consideration as a system for providing low-temperature nuclear heat for district heating.

  8. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    OpenAIRE

    Urva Pancholi; Dhaval Dave; Ajay Patel

    2016-01-01

    The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR), generally of a Canadian Deuterium Uranium (CANDU) type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniqu...

  9. IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

    2004-10-03

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

  10. Effects of breeder age and pre-incubation storage of eggs on ...

    African Journals Online (AJOL)

    S..

    hatched chicks was influenced by the storage period x breeder age ... liver weight as a percentage of live weight was higher when eggs obtained from 10 ... and it remains unclear how these storage conditions affect water vapour loss during.

  11. THE INFLUENCE OF MIEX® RESIN FOR WATER TREATMENT EFFICIENCYIN A HYBRID MEMBRANE REACTOR

    Directory of Open Access Journals (Sweden)

    Mariola Rajca

    2014-10-01

    Full Text Available The paper presents the results of studies related to the effectiveness of removal of natural organic matter (NOM from water using hybrid membrane reactor in which ion exchange and ultrafiltration processes were performed. MIEX® resin by Orica Watercare and immersed ultrafiltration polyvinylidene fluoride capillary module ZeeWeed 1 (ZW 1 by GE Power&Water operated at negative pressure were used. The application of multifunctional reactor had a positive effect on the removal of contaminants and enabled the production of high quality water. Additionally, in refer to single stage ultrafiltration it minimalized the occurrence of membrane fouling.

  12. Irradiation tests of texture controlled cladding for pressurized water reactor in foreign reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ukawa, Kazunori [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Abeta, Sadaaki

    1996-12-01

    PWR electric power companies and makers are promoting a high burn-up program for nuclear fuel aiming at reducing spent fuel. PCI tolerance performance of fuel clad is desirable to promote a high burnup program. It is clear that when we change the direction of a Zircaloy crystal structure by improving a manufacture process (texture control), PCI tolerance performance greatly improved. In this study, an improved clad was burnt in a R2 reactor to 61.5 GWd/t and a power ramp test was carried out in Sweden. Based on irradiation data, power ramp test data and post irradiation examination data, improvement of PCI tolerance performance was confirmed. (author)

  13. Nuclear reactor composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  14. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detector

    Energy Technology Data Exchange (ETDEWEB)

    Hellfeld, D. [Texas A & M Univ., College Station, TX (United States); Dazeley, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bernstein, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marianno, C. [Texas A & M Univ., College Station, TX (United States)

    2015-11-25

    The potential of elastic antineutrino-electron scattering (ν¯e + e → ν¯e + e) in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor. Background was estimated via independent simulations and by appropriately scaling published measurements from similar detectors. Many potential backgrounds were considered, including solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclide and water-borne radon decays, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. The detector response was modeled using a GEANT4-based simulation package. The results indicate that with the use of low radioactivity PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. The directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of theoretical conditions that, if satisfied in practice, would enable nuclear reactor antineutrino directionality in a Gd-doped water Cherenkov detector approximately 10 km from a large power reactor.

  15. Production behavior of irradiation defects in solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Hirotake; Moritani, Kimikazu [Kyoto Univ. (Japan)

    1998-03-01

    The irradiation effects in solid breeder materials are important for the performance assessment of fusion reactor blanket systems. For a clearer understanding of such effects, we have studied the production behavior of irradiation defects in some lithium ceramics by an in-situ luminescence measurement technique under ion beam irradiation. The luminescence spectra were measured at different temperatures, and the temperature-transient behaviors of luminescence intensity were also measured. The production mechanisms of irradiation defects were discussed on the basis of the observations. (author)

  16. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  17. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  18. Inspection and evaluation guidelines for light water reactor internals

    Energy Technology Data Exchange (ETDEWEB)

    Iizuka, N. [Tokyo Electric Power Co., Inc. (Japan); Taniguchi, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Yoshinaga, T. [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    On February, 2000, in the Engineering Society of Thermal and Nuclear Power Generation, the 'Investigation Group on Inspection and Evaluation Guidelines for Nuclear Reactor Internals' was established. This group was started at moments of some damage cases on reactor internals on BWRs and PWRs in Japan and foreign countries and of finding out cracks based on a number of SCC (stress corrosion cracking) at Inconel alloy weldings of a shroud support of BWR internals in Tsuruga Nuclear Power Station Unit-1 of the Japan Nuclear Power Generation Co., Ltd. on December, 1999. Under these conditions, this group made some guidelines for rational inspection with clear technical foundation, and so on as well as arrangements on structural functions, importance at safety, and so on of the reactor internals, promoted some investigations aiming at wide general proposal on how to carry out future internal inspections on LWR in Japan, and completed almost all of the investigations on March, 2002. Here were described basic indications of the guideline development and summaries of the developed guideline. (G.K.)

  19. The ISS Water Processor Catalytic Reactor as a Post Processor for Advanced Water Reclamation Systems

    Science.gov (United States)

    Nalette, Tim; Snowdon, Doug; Pickering, Karen D.; Callahan, Michael

    2007-01-01

    Advanced water processors being developed for NASA s Exploration Initiative rely on phase change technologies and/or biological processes as the primary means of water reclamation. As a result of the phase change, volatile compounds will also be transported into the distillate product stream. The catalytic reactor assembly used in the International Space Station (ISS) water processor assembly, referred to as Volatile Removal Assembly (VRA), has demonstrated high efficiency oxidation of many of these volatile contaminants, such as low molecular weight alcohols and acetic acid, and is considered a viable post treatment system for all advanced water processors. To support this investigation, two ersatz solutions were defined to be used for further evaluation of the VRA. The first solution was developed as part of an internal research and development project at Hamilton Sundstrand (HS) and is based primarily on ISS experience related to the development of the VRA. The second ersatz solution was defined by NASA in support of a study contract to Hamilton Sundstrand to evaluate the VRA as a potential post processor for the Cascade Distillation system being developed by Honeywell. This second ersatz solution contains several low molecular weight alcohols, organic acids, and several inorganic species. A range of residence times, oxygen concentrations and operating temperatures have been studied with both ersatz solutions to provide addition performance capability of the VRA catalyst.

  20. Nonisothermal reactors for the production of pure water from peritoneal dialysis waste waters.

    Science.gov (United States)

    Diano, N; Ettari, G; Grano, V; Gaeta, F S; Rossi, S; Bencivenga, U; D'Alterio, C; Ruocco, G; Mita, L; De Santo, N G; Canciglia, P; Mita, D G

    2007-01-01

    The diffusion of peritoneal dialysis (PD) at home is somewhat restricted by the difficulty of transport and storage of a large amount of dialytic solutions. This problem is exacerbated in the case of hemodialysis. With the aim of producing pure water to be used in preparing the solution for peritoneal dialysis, or for hemodialysis in general, as one example, we purified the spent dialysate solution from PD. Experiments were carried out with 24 dialysate solutions taken from 8 patients. Pure water was obtained by means of a thermodialysis process in a hollow fiber reactor operating under nonisothermal conditions. Results show that the yield of the nonisothermal process is dependent on the temperature difference applied across the hydrophobic membranes. The production of pure water per square meter of membrane and per hour was equal to 0.55 or 1.2 or 2.0 liters, with a temperature difference of 11 degrees C or 21 degrees C or 28 degrees C, respectively. These results encourage the use of the thermodialysis process in the production of pure water for clinical uses.

  1. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  2. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Science.gov (United States)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  3. The TMSR as actinide burner and thorium breeder

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Allibert, M.; Ghetta, V. [LPSC/IN2P3/CNRS - INPG/ENSPG - UJF, 53, avenue des Martyrs, F-38026 Grenoble Cedex (France)

    2007-07-01

    Molten Salt Reactors (MSRs) are one of the six systems retained by Generation IV as a candidate for the next generation of nuclear reactors. Molten Salt Reactor is a very attractive concept especially for the Thorium fuel cycle which allows nuclear energy production with a very low production of radio-toxic minor actinides. Studies have thus been done on the Molten Salt Breeder Reactor (MSBR) of Oak-Ridge to re-evaluate this concept. They have shown that the MSBR suffers from major drawbacks concerning for example safety and reprocessing, drawbacks incompatible with any industrial development. On the other hand, the advantages of the Thorium fuel cycle were too attractive not to look further into it. With these considerations, we have reassessed the whole concept to propose an innovative reactor called Thorium Molten Salt Reactor (TMSR). Many parametric studies of the TMSR have been carried out, correlating the core arrangement and composition, the reprocessing performances, and the salt composition. In particular, by changing the moderation ratio of the core the neutron spectrum can be modified and placed anywhere between a very thermalized neutron spectrum and a relatively fast spectrum. Even if the epithermal TMSR configurations have not been completely excluded by our calculations, our studies have shown that the reactor design where there is no graphite moderator inside the core appears to be the most promising in terms of safety coefficients, reprocessing requirements, and breeding and deployment capabilities. Larger fissile matter inventories are necessary in such a reactor configuration compared to the thermalized TMSR configurations, but the resulting deployment limitation could be solved by using transuranic elements as initial fissile load. This work is based on the coupling of a neutron transport code called MCNP with the materials evolution code REM. The former calculates the neutron flux and the reaction rates in all the cells while the latter solves

  4. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  5. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  6. Component failures at pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

  7. On the study of catalytic membrane reactor for water detritiation: Modeling approach

    Energy Technology Data Exchange (ETDEWEB)

    Liger, Karine, E-mail: karine.liger@cea.fr [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Mascarade, Jérémy [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Joulia, Xavier; Meyer, Xuan-Mi [Université de Toulouse, INPT, UPS, Laboratoire de Génie Chimique, 4, Allée Emile Monso, Toulouse F-31030 (France); CNRS, Laboratoire de Génie Chimique, Toulouse F-31030 (France); Troulay, Michèle; Perrais, Christophe [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France)

    2016-11-01

    Highlights: • Experimental results for the conversion of tritiated water (using deuterium as a simulant of tritium) by means of a catalytic membrane reactor in view of tritium recovery. • Phenomenological 2D model to represent catalytic membrane reactor behavior including the determination of the compositions of gaseous effluents. • Good agreement between the simulation results and experimental measurements performed on the dedicated facility. • Explanation of the unexpected behavior of the catalytic membrane reactor by the modeling results and in particular the gas composition estimation. - Abstract: In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium bonded water can be recovered under the valuable Q{sub 2} form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal of reactions products all along the reactor, and thus limits reverse reaction rate to the benefit of the direct one (shift effect). The reactions kinetics, which are still little known or unknown, are generally assumed to be largely greater than the permeation ones so that thermodynamic equilibriums of isotope exchange reactions are generally assumed. This paper proposes a new phenomenological 2D model to represent catalytic membrane reactor behavior with the determination of gas effluents compositions. A good agreement was obtained between the simulation results and experimental measurements performed on a dedicated facility. Furthermore, the gas composition estimation permits to interpret unexpected behavior of the catalytic membrane reactor. In the next future, further sensitivity analysis will be performed to determine the limits of the model and a kinetics study will be conducted to assess the thermodynamic equilibrium of reactions.

  8. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  9. Construction of the advanced boiling water reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan). Nuclear Power Plant Construction Dept.

    1996-07-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7.

  10. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

  11. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Diggs, J M

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables.

  12. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    Science.gov (United States)

    Hellfeld, D.; Bernstein, A.; Dazeley, S.; Marianno, C.

    2017-01-01

    The potential of elastic antineutrino-electron scattering in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13-km standoff from a 3.758-GWt light water nuclear reactor and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays from the photomultiplier tubes (PMTs), detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. The results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3σ significance in large Gd-doped water Cherenkov detectors with greater than 10-km standoff from a nuclear reactor.

  13. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  14. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [and others

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  15. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  16. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [Argonne National Lab., IL (United States)] [and others

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  17. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gavilian-Moreno, Carlos [Iberdrola Generacion, S.A., Cofrentes Nuclear Power Plant, Project Engineering Department, Paraje le Plano S/N, Valencia (Spain); Espinosa-Paredes, Gilberto [Area de ingeniera en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Mexico city (Mexico)

    2016-04-15

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  18. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  19. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mausolf, Edward J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mcnamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schwantes, Jon M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  20. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  1. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  2. Hydraulic performance of a proposed in situ photocatalytic reactor for degradation of MTBE in water.

    Science.gov (United States)

    Lim, Leonard Lik Pueh; Lynch, Rod

    2011-01-01

    Methyl tert-butyl ether (MTBE) groundwater remediation projects often require a combination of technologies resulting in increasing the project costs. A cost-effective in situ photocatalytic reactor design, Honeycomb II, is proposed and tested for its efficiency in MTBE degradation at various flows. This study is an intermediate phase of the research in developing an in situ photocatalytic reactor for groundwater remediation. It examines the effect of the operating variables: air and water flow and double passages through Honeycomb II, on the MTBE removal. MTBE vaporisation is affected by not only temperature, Henry's law constant and air flow to volume ratio but also reactor geometry. The column reactor achieved more than 84% MTBE removal after 8 h at flows equivalent to horizontal groundwater velocities slower than 21.2 cm d⁻¹. Despite the contrasting properties between a photocatalytic indicator methylene blue and MTBE, the reactor efficiency in degrading both compounds showed similar responses towards flow (equivalent groundwater velocity and hydraulic residence time (HRT)). The critical HRT for both compounds was approximately 1 d, which corresponded to a velocity of 21.2 cm d⁻¹. A double pass through both new and used catalysts achieved more than 95% MTBE removal after two passes in 48 h. It also verified that the removal efficiency can be estimated via the sequential order of the removal efficiency of one pass obtained in the laboratory. This study reinforces the potential of this reactor design for in situ groundwater remediation.

  3. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  4. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    Energy Technology Data Exchange (ETDEWEB)

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  5. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

  6. Breeder nutrition and offspring performance

    Directory of Open Access Journals (Sweden)

    F Calini

    2007-06-01

    Full Text Available Vertical integration in poultry industry strongly emphasizes the importance of cost control at all levels. In the usual broiler production operations, the costs involved with the production of the hatching egg or the day old chick are negligible if seen in the perspective of the cost per kg of live bird. From a research point of view, anyway, the greatest attention is usually given to the performance of broiler breeders, and most of the research in the field is focused on the improvement of their relative performance, mainly in terms of saleable chicks produced per hen, while less attention has been given to the quality of the chick and to the improvement of its growth performances, even if these last parameters have an effective impact on the overall economics of the poultry growing business. Most of the data available is quite dated, as can be seen from some recent reviews, and in general little attention is given to the impact of parental nutrition on the subsequent broiler performance. It is in fact more usual to find data about dam nutrition influence on egg fertility and hatchability than on subsequent progeny performance. The objectives of this review were to assess, on the basis of published reports, the effects of selected nutrients and anti-nutrients normally prevailing in commercial broiler breeder feeds - vitamins, micro-minerals, mycotoxins, - trying to pinpoint which could be the positive and the negative effects of both on the subsequent broiler performance, with a particular attention to the impact on immune function and carcass yield.

  7. TMI-2 Reactor Building source term measurements: surfaces and basement water and sediment

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C V; Keefer, D G

    1984-10-01

    Presented in this report are the results of radiochemical and elemental analyses performed on samples collected from the Three Mile Island Unit 2 Reactor Building from August 1979 to December 1983. The quantities of fission products and core materials that were measured on the external surfaces in the Reactor Building or in the water and sediment in its basement are summarized. Recent analysis results for access panels removed from the air cooling assembly and for liquid and particulate samples collected from the Reactor Building sump and reactor coolant drain tank are included in the report. Measurements show that 59% of the /sup 3/H, 2.7% of the /sup 90/Sr, 15% of the /sup 129/I, 20% of the /sup 131/I, and 42% of the /sup 137/Cs originally in the core at the time of the accident could be accounted for outside the core in the Reactor Building. With the exceptions of /sup 90/Sr and /sup 144/Ce, the vast majority of each radionuclide released was found dispersed in the water and sediment in the basement.

  8. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  9. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

  10. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition, numerica

  11. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures...

  12. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition, numerica

  13. Neutron fluence depth profiles in water phantom on epithermal beam of LVR-15 research reactor.

    Science.gov (United States)

    Viererbl, L; Klupak, V; Lahodova, Z; Marek, M; Burian, J

    2010-01-01

    Horizontal channel with epithermal neutron beam at the LVR-15 research reactor is used mainly for boron neutron capture therapy. Neutron fluence depth profiles in a water phantom characterise beam properties. The neutron fluence (approximated by reaction rates) depth profiles were measured with six different types of activation detectors. The profiles were determined for thermal, epithermal and fast neutrons.

  14. Water plant modifications for increased production at B, C, D, DR, F, and H Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, L.B.; Corley, J.P.

    1960-04-15

    The purpose of this report is to define the extent of modifications necessary to increase capacities of the 100-B, C, D, DR, F, and H water plants for reactor flows of 90,000 95,000 105,000 and 115,000 GPM, and to provide supporting data for budget studies for increased production.

  15. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition,

  16. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security... Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical output... from the regulations in part 73 as it determines are authorized by law and will not endanger life...

  17. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  18. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  19. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

  20. Design characteristics for pressurized water small modular nuclear power reactors with focus on safety

    Energy Technology Data Exchange (ETDEWEB)

    Kani, Iraj Mahmoudzadeh [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; Zandieh, Mehdi [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty; Abadi, Saeed Kheirollahi Hossein [International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty

    2016-05-15

    Small Modular Reactors (SMRs) are a technology, attracting attention. Light water SMR possess an upgraded design case and emphasize the significance of integral models. Beside of these advantages, SMRs has faced numerous challenges, e.g. licensing, cost/investment, safety and security observation, social and environmental issues in building new plants.

  1. Environmentally assisted cracking in light water reactors. Semiannual report, April--September 1991: Volume 13

    Energy Technology Data Exchange (ETDEWEB)

    Kassner, T F; Ruther, W E; Chung, H M; Hicks, P D; Hins, A G; Park, J Y; Soppet, W K; Shack, W J [Argonne National Lab., IL (United States)

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with {approx} 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  2. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Ho; Kim, Dae Seop; Kim, Jae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2014-06-15

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

  3. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    Science.gov (United States)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  4. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

  5. Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor.

    Science.gov (United States)

    Polo-López, M I; Fernández-Ibáñez, P; Ubomba-Jaswa, E; Navntoft, C; García-Fernández, I; Dunlop, P S M; Schmid, M; Byrne, J A; McGuigan, K G

    2011-11-30

    Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

  6. Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor

    Energy Technology Data Exchange (ETDEWEB)

    Polo-Lopez, M.I., E-mail: mpolo@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Fernandez-Ibanez, P., E-mail: pilar.fernandez@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Ubomba-Jaswa, E., E-mail: euniceubombajaswa@yahoo.com [Natural Resources and the Environment, CSIR, PO Box 395, Pretoria (South Africa); Navntoft, C., E-mail: christian.navntoft@solarmate.com.ar [Instituto de Investigacion e Ingenieria Ambiental, Universidad Nacional de San Martin (3iA-UNSAM), Peatonal Belgrano 3563, B1650ANQ San Martin (Argentina); Universidad Tecnologica Nacional - Facultad Regional Buenos Aires - Departamento de Ingenieria Civil - Laboratorio de Estudios sobre Energia Solar, (UTN-FRBA-LESES), Mozart 2300, (1407) Ciudad Autonoma de Buenos Aires, Republica Argentina (Argentina); Garcia-Fernandez, I., E-mail: irene.garcia@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Dunlop, P.S.M., E-mail: psm.dunlop@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Schmid, M. [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Byrne, J.A., E-mail: j.byrne@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); and others

    2011-11-30

    Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

  7. Non normal modal analysis of oscillations in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas

    2013-07-01

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

  8. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  9. Experimental techniques to determine salt formation and deposition in supercritical water oxidation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chan, J.P.C.; LaJeunesse, C.A.; Rice, S.F.

    1994-08-01

    Supercritical Water Oxidation (SCWO) is an emerging technology for destroying aqueous organic waste. Feed material, containing organic waste at concentrations typically less than 10 wt % in water, is pressurized and heated to conditions above water`s critical point where the ability of water to dissolve hydrocarbons and other organic chemicals is greatly enhanced. An oxidizer, is then added to the feed. Given adequate residence time and reaction temperature, the SCWO process rapidly produces innocuous combustion products. Organic carbon and nitrogen in the feed emerge as CO{sub 2} and N{sub 2}; metals, heteroatoms, and halides appear in the effluent as inorganic salts and acids. The oxidation of organic material containing heteroatoms, such as sulfur or phosphorous, forms acid anions. In the presence of metal ions, salts are formed and precipitate out of the supercritical fluid. In a tubular configured reactor, these salts agglomerate, adhere to the reactor wall, and eventually interfere by causing a flow restriction in the reactor leading to an increase in pressure. This rapid precipitation is due to an extreme drop in salt solubility that occurs as the feed stream becomes supercritical. To design a system that can accommodate the formation of these salts, it is important to understand the deposition process quantitatively. A phenomenological model is developed in this paper to predict the time that reactor pressure begins to rise as a function of the fluid axial temperature profile and effective solubility curve. The experimental techniques used to generate effective solubility curves for one salt of interest, Na{sub 2}SO{sub 4}, are described, and data is generated for comparison. Good correlation between the model and experiment is shown. An operational technique is also discussed that allows the deposited salt to be redissolved in a single phase and removed from the affected portion of the reactor. This technique is demonstrated experimentally.

  10. The controllability analysis of the purification system for heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. D.; Cho, B. H.; Shin, C. H.; Kim, S. H. [KEPRI, Taejon (Korea, Republic of); Lee, Y. K.; Kim, K. U. [KHNP, Kyungju (Korea, Republic of)

    2001-10-01

    The heavy water reactor such as Wolsung No.1 and No.2 has a purification system to purify the reactor coolant. The control system regulates the coolant temperature to protect the ion exchanger. After the fuel exchanges of operating plant, the increase of the coolant pressure makes the purification temperature control difficult. In this paper, the controllability of the control dynamics of the purification system was analysed and the optimal parameters were proposed. To reduce the effects of the flow disturbance, the feedforward control structure was proposed and analysed.

  11. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    Directory of Open Access Journals (Sweden)

    Urva Pancholi

    2016-07-01

    Full Text Available The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR, generally of a Canadian Deuterium Uranium (CANDU type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniques, instruments and specific data for such components mentioned here are synthesized results from primary research and knowledge available in this area, in order to produce coherent argument focused on quality control of end fittings.

  12. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    Energy Technology Data Exchange (ETDEWEB)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

  13. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    Energy Technology Data Exchange (ETDEWEB)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330° and 370°C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  14. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    Energy Technology Data Exchange (ETDEWEB)

    Luscher, Walter G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Senor, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Kevin K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Longhurst, Glen R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ºC). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  15. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W [Westinghouse Savannah River Co., Aiken, SC (USA); Hagrman, D L [EG and G Idaho, Inc., Idaho Falls, ID (USA); McClure, P R; Leonard, M T [Science Applications International Corp., Albuquerque, NM (USA)

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  16. Degraded core analysis for the pressurized-water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-02-09

    An analysis of the likelihood and the consequences of 'degraded-core accidents' has been undertaken for the proposed Sizewell B PWR. In such accidents, degradation of the core geometry occurs as a result of overheating. Radionuclides are released and may enter the environment, causing harmful effects. The analysis concludes that degraded-core accidents are highly improbable, the plant having been designed to reduce the frequency of such accidents to a value of order 10/sup -6/ per year. Tbe building containing the reactor would only fail in a small proportion of degraded-core accidents. In the great majority of cases the containment would remain intact and the release of radioactivity to the environment would be small. The risk to individuals have been calculated for both immediate and long term effects. Although the estimates of risk are approximate, studies to investigate the uncertainties, and sensitivities to different assumptions, show that potential errors are small compared with the very large 'margin of safety' between the risks estimated for Sizewell B and those that already exist in society.

  17. Biological mine water treatment operating a one stage reactor system

    CSIR Research Space (South Africa)

    Baloyi, MJ

    2006-05-01

    Full Text Available Mine drainage arises from oxidation of pyrites, due to exposure to air and water. Acid mine drainage normally contains high concentrations of sulphate, metals and acidity. These pollutants can be reduced by applying the biological sulphate reduction...

  18. Structural analysis of reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design.

  19. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-10-03

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in ancircular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mas velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel.

  20. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  1. [Radiation ecological environment in the Republic of Kazakhstan in the vicinity of the reactors and on the territory of the Semipalatinsk Test Site].

    Science.gov (United States)

    Kim, D S

    2012-01-01

    The results of research into the environmental conditions in the regions of location of the pressurized water reactor WWR-K, fast neutron breeder BN-350 and on the territory of the Semipalatinsk Test Site are represented. The effects of the exposure to aerosol emissions from WWR-K and BN-350 reactors on the environment are summarized. We present some arguments in favor of the safe operation of fission reactors in compliance with the rules and norms of nuclear and radiation protection and the efficient disposal of radioactive waste on the territory of the Republic.

  2. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  3. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  4. The D&D of the Experimental Boiling Water Reactor (EBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Boling, L.E.; Yule, T.J.; Bhattacharyya, S.K.

    1996-03-01

    Argonne National Laboratory has completed the D&D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D&D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort.

  5. Photolytic treatment of atrazine-contaminated water: products, kinetics, and reactor design.

    Science.gov (United States)

    Ye, Xuejun; Chen, Daniel; Li, Kuyen; Wang, Bin; Hopper, Jack

    2007-08-01

    This study investigates the products, kinetics, and reactor design of atrazine photolysis under 254-nm ultraviolet-C (UVC) irradiation. With an initial atrazine concentration of 60 microg/L (60 ppbm), only two products remain in detectable levels. Up to 77% of decomposed atrazine becomes hydroxyatrazine, the major product. Both atrazine and hydroxyatrazine photodecompose following the first-order rate equation, but the hydroxyatrazine photodecomposition rate is significantly slower than that of atrazine. For atrazine photodecomposition, the rate constant is proportional to the square of UVC output, but inversely proportional to the reactor volume. For a photochemical reactor design, a series of equations are proposed to calculate the needed UVC output power, water treatment capacity, and atrazine outlet concentration.

  6. Production aspects of broiler breeders submitted to different drinker types

    Directory of Open Access Journals (Sweden)

    LP Colvero

    2014-03-01

    Full Text Available The objective of this study was of evaluate the influence of different drinker types on the egg production, water intake, mortality, poultry litter relative humidity, egg weight, eggshell percentage, and egg specific gravity of broiler breeders. The experiment was carried out in a commercial farm with 37- to 44-wk-old broiler breeders. A randomized block experimental design, consisting of two treatments (bell or nipple drinkers with four replicates of 4.000 females each, was applied. Data were submitted to analysis of variance, and means were compared by the test of Student-Newman-Keuls at 5% significance level. Birds submitted to nipple drinkers presented lower water intake (p0.05 of drinker type on egg production or mortality. Poultry litter relative humidity was lower (p<0.05 under the nipple-drinker system. Birds drinking from bell drinkers produced heavier eggs (p<0.05 between weeks 39 and 40. Hens drinking from bell drinkers laid eggs with higher specific gravity and eggshell percentage. It was concluded that nipple drinkers can be used for broiler breeders during lay.

  7. Drinking water treatment in solar reactors with immobilized photocatalysts

    Energy Technology Data Exchange (ETDEWEB)

    Sichel, C.; Fernandez, P.; Blanco, J.; Lorenz, K.

    2005-07-01

    This work has been performed at the Plataforma Solar de Almeria. As in our daily consumption of any other good, it is important to take an interest in sustainable treatment methods for purifying a vital water supply. Primary water treatment has no need for energy consuming techniques as any suspended particles can usually be removed by sand traps and sedimentation basin. Organic matter and biodegradable chemical contaminants ca be decomposed by activated sludge plants, bacteria beds, or in the case of highly organically loaded sewage by methanisation.In the recent years, another photocatalysts a photo sensitizer has been used in desinfection experiments. Ruthenium appears to have good potential for inactivation of bacteria in chelating coordination compounds. The SOLWATER project attempts to provide remote areas of such developing countries as Mexico, Peru and Argentina with drinking water disinfected by solar photocatalysis with immobilized TiO2 and Ru(II). (Author)

  8. Laboratory data for review of outlet water temperature limits for BDF type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.; Fitzsimmons, D.E.

    1960-12-11

    A knowledge of the thermal and hydraulic conditions within a reactor fuel channel during an inadvertent flow reduction is needed to establish reactor operating limits. Such limits, which are based on outlet water temperature, are called ``trip-after-instability`` limits by the reactor operating personnel. Laboratory experiments were performed to update the knowledge of such conditions in a BDF reactor type fuel channel while using internally and externally cooled fuel elements (I&E`s) at tube powers up to 1530 KW. In addition to a general extension of previous data, the new data were used to review certain specific details involved in ``trip-after-instability`` limit calculations. It was found that in calculating the limits, the isothermal pressure drop across the fuel elements must be related to flow rate by the exponent 1.8, ({delta}P {proportional_to} F{sup 1.8}), rather than the more convenient value of 2.0. It was found that this method of limit determination is applicable to the high rear header pressures presently attained on the reactors and also applicable to tubes with very low Panellit pressures. And finally, the validity of certain analytical transformations of experimental data, called generalization of hydraulic demand curves, was reaffirmed for the above conditions.

  9. Microfluidic reactors for visible-light photocatalytic water purification assisted with thermolysis.

    Science.gov (United States)

    Wang, Ning; Tan, Furui; Wan, Li; Wu, Mengchun; Zhang, Xuming

    2014-09-01

    Photocatalytic water purification using visible light is under intense research in the hope to use sunlight efficiently, but the conventional bulk reactors are slow and complicated. This paper presents an integrated microfluidic planar reactor for visible-light photocatalysis with the merits of fine flow control, short reaction time, small sample volume, and long photocatalyst durability. One additional feature is that it enables one to use both the light and the heat energy of the light source simultaneously. The reactor consists of a BiVO4-coated glass as the substrate, a blank glass slide as the cover, and a UV-curable adhesive layer as the spacer and sealant. A blue light emitting diode panel (footprint 10 mm × 10 mm) is mounted on the microreactor to provide uniform irradiation over the whole reactor chamber, ensuring optimal utilization of the photons and easy adjustments of the light intensity and the reaction temperature. This microreactor may provide a versatile platform for studying the photocatalysis under combined conditions such as different temperatures, different light intensities, and different flow rates. Moreover, the microreactor demonstrates significant photodegradation with a reaction time of about 10 s, much shorter than typically a few hours using the bulk reactors, showing its potential as a rapid kit for characterization of photocatalyst performance.

  10. Single-stage temperature-controllable water gas shift reactor with catalytic nickel plates

    Science.gov (United States)

    Park, Jin-Woo; Lee, Sung-Wook; Lee, Chun-Boo; Park, Jong-Soo; Lee, Dong-Wook; Kim, Sung-Hyun; Kim, Sung-Soo; Ryi, Shin-Kun

    2014-02-01

    In this study, a microstructured reactor with catalytic nickel plates is newly designed and developed for proper heat management in an exothermic water gas shift WGS reaction. The reactor is designed to increase the reactor capacity simply by numbering-up a set of a catalyst layers and heat exchanger layers. The WGS reactor is built up with two sets of a catalyst layers and heat exchanger layers. The performance of the reactor is verified by WGS testing with the variation of the furnace temperatures, gas hourly space velocity (GHSV) and coolant (N2) flow rate. At a GHSV of 10,000 h-1, CO conversion reaches the equilibrium value with a CH4 selectivity of ≤0.5% at the furnace temperature of ≥375 °C. At high GHSV (40,000 h-1), CO conversion decreases considerably because of the heat from the exothermic WGS reaction at a large reactants mass. By increasing the coolant flow rate, the heat from the WGS reaction is properly managed, leading an increase of the CO conversion to the equilibrium value at GHSV of 40,000 h-1.

  11. Corrosion mechanisms of candidate structural materials for supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Lefu ZHANG; Fawen ZHU; Rui TANG

    2009-01-01

    Nickel-based alloys, austenitic stainless steel, ferritic/martensitic heat-resistant steels, and oxide dispersion strengthened steel are presently considered to be the candidate structural or fuel-cladding materials for supercritical water-cooled reactor (SCWR), one of the promising generation IV reactor for large-scale electric power production. However, corrosion and stress corrosion cracking of these candidate alloys still remain to be a major problem in the selection of nuclear fuel cladding and other structural materials, such as water rod. Survey of literature and experimental results reveal that the general corrosion mechanism of those candidate materials exhibits quite complicated mechanism in high-temperature and high-pressure supercritical water. Formation of a stable protective oxide film is the key to the best corrosion-resistant alloys. This paper focuses on the mechanism of corrosion oxide film breakdown for SCWR candidate materials.

  12. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  13. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson J. Boise; Reid, Robert S.

    2007-01-01

    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  14. Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  15. Space reactors. Progress report, July-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Ranken, W.A. (comp.)

    1982-06-01

    Progress in technology development for the Space Power Advanced Reactor (SPAR) project is reported for the period July 1-September 30, 1981. The weights of neutron and gamma shields required to protect the SPAR system payloads for a range of permissible exposures have been determined, and initial results are reported. SPAR reactor safety in the case of water immersion has been modeled. Approval-in-Principle has been received for the SPAR fuel test in the Experimental Breeder Reactor-II (EBR-II); the heat pipe developed for this test is performing well. SPAR system design variations are being examined under the possibility of using long core heat pipes. Testing of the initial molybdenum/sodium artery heat pipe continued, with ambiguous results. Fabrication of the first all-bonded thermoelectric units has been completed and testing has been initiated.

  16. Gas phase photocatalytic water splitting in silicon based µ-reactors

    DEFF Research Database (Denmark)

    Dionigi, Fabio; Vesborg, Peter Christian Kjærgaard

    to the water splitting experiments, the results obtained with SrTiO2 and TiO2 are presented. These semiconductors are well known examples of materials active under UV illumination. However to achieve high efficiency of solar energy conversion the catalysts needs to be active for longer wavelength. Ga......N:ZnO is one of the few photocatalysts that is able to achieve overall water splitting with visible light. Therefore the reaction has been studied focusing on this material. GaN:ZnO loaded with Rh2-yCryO3 showed high activity and hydrogen and oxygen could even be detected under illumination with a solar light...... to the products detection using μ-reactors. In particular a new kind of μ-reactor that has a Pyrex lid on both sides is presented. With this reactor is possible to measure the absorbance of the materials deposited inside the μ-reactor and to combine optical measurements and spectroscopy with the detection...

  17. Biomimetic microchannels of planar reactors for optimized photocatalytic efficiency of water purification

    Science.gov (United States)

    Liao, Wuxia; Wang, Ning; Wang, Taisheng; Xu, Jia; Han, Xudong; Liu, Zhenyu; Yu, Weixing

    2016-01-01

    This paper reports a biomimetic design of microchannels in the planar reactors with the aim to optimize the photocatalytic efficiency of water purification. Inspired from biology, a bifurcated microchannel has been designed based on the Murray's law to connect to the reaction chamber for photocatalytic reaction. The microchannels are designed to have a constant depth of 50 μm but variable aspect ratios ranging from 0.015 to 0.125. To prove its effectiveness for photocatalytic water purification, the biomimetic planar reactors have been tested and compared with the non-biomimetic ones, showing an improvement of the degradation efficiency by 68%. By employing the finite element method, the flow process of the designed microchannel reactors has been simulated and analyzed. It is found that the biomimetic design owns a larger flow velocity fluctuation than that of the non-biomimetic one, which in turn results in a faster photocatalytic reaction speed. Such a biomimetic design paves the way for the design of more efficient planar reactors and may also find applications in other microfluidic systems that involve the use of microchannels. PMID:26958102

  18. A Study on the Optimal Position for the Secondary Neutron Source in Pressurized Water Reactors

    Directory of Open Access Journals (Sweden)

    Jungwon Sun

    2016-12-01

    Full Text Available This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

  19. Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

    Science.gov (United States)

    Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

    2006-06-01

    Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

  20. Reactor Materials Program process water piping indirect failure frequency

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.L.

    1989-10-30

    Following completion of the probabilistic analyses, the LOCA Definition Project has been subject to various external reviews, and as a result the need for several revisions has arisen. This report updates and summarizes the indirect failure frequency analysis for the process water piping. In this report, a conservatism of the earlier analysis is removed, supporting lower failure frequency estimates. The analysis results are also reinterpreted in light of subsequent review comments.

  1. A Management Strategy for the Heavy Water Reflector Cooling System of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H. S.; Park, Y. C.; Lim, S. P. (and others)

    2007-11-15

    Heavy water is used as the reflector and the moderator of the HANARO research reactor. After over 10 years operation since first criticality in 1995 there arose some operational issues related with the tritium. A task force team(TFT) has been operated for 1 year since September 2006 to study and deduce resolutions of the issues concerning the tritium and the degradation of heavy water in the HANARO reflector system. The TFT drew many recommendations on the hardware upgrade, tritium containing air control, heavy water quality management, waste management, and tritium measurement system upgrade.

  2. Efficiency Calibration of LaBr3(Ce) γ Spectroscopy in Analyzing Radionucles in Reactor Loop Water

    Institute of Scientific and Technical Information of China (English)

    CHEN; Xi-lin; QIN; Guo-xiu; GUO; Xiao-qing; CHEN; Yong-yong; MENG; Jun

    2013-01-01

    Monitoring the occurring and radioactivity concentration of fission products in nuclear reactor loop water is important for the nuclear reactor safe running evaluation,prevention of accidence and safe protection of working personnel.Study on the efficiency calibration for a LaBr3(Ce)detector experimental

  3. Boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tokarev, Yu.I.; Sokolov, I.N.; Skvortsov, S.A.; Sidorov, A.M.; Krauze, L.V.

    1978-04-01

    The possibility of using a boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant (CHPP) was considered, with design features of the reactor intended for a two-purpose plant. A prestressed reinforced concrete vessel and integral arrangement of the primary circuit ensured reliability of the atomic CHPP using various CHPP flowsheets.

  4. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  5. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Science.gov (United States)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  6. Structural Integrity of Water Reactor Pressure Boundary Components.

    Science.gov (United States)

    1981-02-20

    RES-79-103 UNCLASSIFIED NRL--- 400 NURE-CR-17B3 NL mnmmnuunin -’El-.--. IIIIIIINI ., *q. - - ,aM T? * NUREG /CI 73 NIL Iteof AW, SOIituA 1 nert of Water...Progress Report for July-September 1979," NUREG /CR-1197, Oak Ridge National Labora- tory, Oak Ridge, Tn., Oct. 1978. 2. F. J. Loss, Ed., "Structural...Progress Report for April-June 1976," ORNL/ NUREG /TM-49, Oak Ridge National Labora- tory, Oak Ridge, Tn., Oct. 1976, pp. 27-38. 5. R. G. Berggren

  7. Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mohora, Emilijan, E-mail: emohora@ifc.org [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia); Roncevic, Srdjan; Dalmacija, Bozo; Agbaba, Jasmina; Watson, Malcolm; Karlovic, Elvira; Dalmacija, Milena [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A continuous electrocoagulation/flotation reactor was designed built and operated. Black-Right-Pointing-Pointer Highest NOM removal according to UV{sub 254} was 77% relative to raw groundwater. Black-Right-Pointing-Pointer Highest NOM removal accordance to DOC was 71%, relative to raw groundwater. Black-Right-Pointing-Pointer Highest As removal archived was 85% (6.2 {mu}g/l), relative to raw groundwater. Black-Right-Pointing-Pointer Specific reactor energy and electrode consumption was 1.7 kWh/m{sup 3} and 66 g Al/m{sup 3}. - Abstract: The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate = 4.3 l/h, inter electrode distance = 2.8 cm, current density = 5.78 mA/cm{sup 2}, A/V ratio = 0.248 cm{sup -1}. The NOM removal according to UV{sub 254} absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 {mu}g As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m{sup 3}. According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater.

  8. Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.; Gaitonde, S.M.

    1982-09-01

    The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

  9. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  10. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

    Directory of Open Access Journals (Sweden)

    Diego Mandelli

    2015-01-01

    Full Text Available In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

  11. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    Science.gov (United States)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  12. Removal of dissolved oxygen from water using a Pd-resin based catalytic reactor

    Institute of Scientific and Technical Information of China (English)

    Wenxin SHI; Chongwei CUI; Liye ZHAO; Shuili YU; Xia YUN

    2009-01-01

    The removal of dissolved oxygen (DO) from water was studied experimentally in a Pd-resin base catalyst reactor using purified hydrogen gas as a reducing agent. The effects of various operating conditions, such as hydrogen and water flow rates, height of the catalytic resin bed, temperature, pH value and nan time, on the removal of DO, had been studied extensively. The results shows that DO could be removed by the reactor from ppm to ppb levels at ambient temperature. Increases of temperature, H2gas rate and the height of the catalytic resin were helpful to improve the DO removal rate. The change of pH value fom 4 to 12 resulted in no effect on DO removal. Reaction time was the key factor to control the DO removal efficiency. Only when the reaction time was longer than 2.3 minutes under the experimental conditions, could a very low DO level be achieved.

  13. Flow-induced vibration for light water reactors. Final progress report, July 1981-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Torres, M.R.

    1981-11-01

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the final period from July 1981 to September 1981. This is the last quarterly progress report to be issued for this program.

  14. Molecular Dynamics Simulations of Aqueous and Confined Systems Relevant to the Supercritical Water Cooled Nuclear Reactor

    Science.gov (United States)

    Kallikragas, Dimitrios Theofanis

    Supercritical water (SCW) is the intended heat transfer fluid and potential neutron moderator in the proposed GEN-IV Supercritical Water Cooled Reactor (SCWR). The oxidative environment poses challenges in choosing appropriate design materials, and the behaviour of SCW within crevices of the passivation layer is needed for developing a corrosion control strategy to minimize corrosion. Molecular Dynamics simulations have been employed to obtain diffusion coefficients, coordination number and surface density characteristics, of water and chloride in nanometer-spaced iron hydroxide surfaces. Diffusion models for hydrazine are evaluated along with hydration data. Results demonstrate that water is more likely to accumulate on the surface at low density conditions. The effect of confinement on the water structure diminishes as the gap size increases. The diffusion coefficient of chloride decreases with larger surface spacing. Clustering of water at the surface implies that the SCWR will be most susceptible to pitting corrosion and stress corrosion cracking.

  15. Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, A.R.; Judkins, R.R. (comps.)

    1979-12-01

    The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O/sub 2/ fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required.

  16. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  17. Computational fluid dynamic analysis of a closure head penetration in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, D.R.; Schwirian, R.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-09-01

    ALLOY 600 has been used typically for penetrations through the closure head in pressurized water reactors because of its thermal compatibility with carbon steel, superior resistance to chloride attack and higher strength than the austenitic stainless steels. Recent plant operating experience with this alloy has indicated that this material may be susceptible to degradation. One of the major parameters relating to degradation of the head penetrations are the operational temperatures and stress levels in the penetration.

  18. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  19. The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, C.-W., E-mail: d937121@oz.nthu.edu.t [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Wang, J.-R.; Lin, H.-T. [Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Cheng, S.-C. [Department of Nuclear Engineering, Taiwan Power Company, No. 242, Sec. 3, Roosevelt Rd., Taipei City 10016, Taiwan (China)

    2010-10-15

    A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using

  20. Risk assessment of a pressurized water reactor for Class 3-8 accidents

    Energy Technology Data Exchange (ETDEWEB)

    Hall, R.E.

    1979-10-01

    An assessment has been made of the impact on societal risk of Class 3-8 accident sequences as defined by Appendix D to 10 CFR50. The present analysis concentrates on a pressurized water reactor and utilizes realistic assumptions when practical. Conclusions are drawn as to the relative importance of the analyzed accidents and their impact on the development of a complete societal risk curve. 65 refs., 61 figs., 37 tabs.

  1. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    OpenAIRE

    Masaki Kurata

    2016-01-01

    Research and development (R&D) methodology for the practical use of accident tolerant fuel (ATF) in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening proces...

  2. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    OpenAIRE

    Hyun-Gil Kim; Jae-Ho Yang; Weon-Ju Kim; Yang-Hyun Koo

    2016-01-01

    For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density compo...

  3. WATER-GAS SHIFT KINETICS OVER IRON OXIDE CATALYSTS AT MEMBRANE REACTOR CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Carl R.F. Lund

    2002-08-02

    The kinetics of water-gas shift were studied over ferrochrome catalysts under conditions with high carbon dioxide partial pressures, such as would be expected in a membrane reactor. The catalyst activity is inhibited by increasing carbon dioxide partial pressure. A microkinetic model of the reaction kinetics was developed. The model indicated that catalyst performance could be improved by decreasing the strength of surface oxygen bonds. Literature data indicated that adding either ceria or copper to the catalyst as a promoter might impart this desired effect. Ceria-promoted ferrochrome catalysts did not perform any better than unpromoted catalyst at the conditions tested, but copper-promoted ferrochrome catalysts did offer an improvement over the base ferrochrome material. A different class of water-gas shift catalyst, sulfided CoMo/Al{sub 2}O{sub 3} is not affected by carbon dioxide and may be a good alternative to the ferrochrome system, provided other constraints, notably the requisite sulfur level and maximum temperature, are not too limiting. A model was developed for an adiabatic, high-temperature water-gas shift membrane reactor. Simulation results indicate that an excess of steam in the feed (three moles of water per mole of CO) is beneficial even in a membrane reactor as it reduces the rate of adiabatic temperature rise. The simulations also indicate that much greater improvement can be attained by improving the catalyst as opposed to improving the membrane. Further, eliminating the inhibition by carbon dioxide will have a greater impact than will increasing the catalyst activity (assuming inhibition is still operative). Follow-up research into the use of sulfide catalysts with continued kinetic and reactor modeling is suggested.

  4. Scram simulation of a control rod drive mechanism of a pressurized water reactor under seismic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, Katsuhisa; Shinohara, Yoshikazu; Ichinoo, Hiroyuki; Yoshikawa, Eiji; Nambu, Kiyoshi; Nomura, Tomonori.

    1987-03-01

    Control rod drop verification experiments of Mitsubishi pressurized water reactor under seismic conditions are performed to confirm the insertion function of control rods into the core. To evaluate these tests, computer simulations are performed. The scram time of control rods under seismic conditions was confirmed to meet the scram function. The behavior of the dropping control rods and the scram time obtained by the computer simulation show a very good correspondence with the results of verification experiments.

  5. On the factors influencing the performance of solar reactors for water disinfection with photosensitized singlet oxygen.

    Science.gov (United States)

    Manjón, Francisco; Villén, Laura; García-Fresnadillo, David; Orellana, Guillermo

    2008-01-01

    Two solar reactors based on compound parabolic collectors (CPCs) were optimized for water disinfection by photosensitized singlet oxygen (1O2) production in the heterogeneous phase. Sensitizing materials containing Ru(II) complexes immobilized on porous silicone were produced, photochemically characterized, and successfully tested for the inactivation of up to 10(4) CFU mL(-1) of waterborne Escherichia coli (gram-negative) or Enterococcus faecalis (gram-positive) bacteria. The main factors determining the performance of the solar reactors are the type of photosensitizing material, the sensitizer loading, the CPC collector geometry (fin- vs coaxial-type), the fluid rheology, and the balance between concurrent photothermal--photolytic and 1O2 effects on the microorganisms' inactivation. In this way, at the 40 degrees N latitude of Spain, water can be disinfected on a sunny day (0.6-0.8 MJ m(-2) L(-1) accumulated solar radiation dose in the 360-700 nm range, typically 5-6 h of sunlight) with a fin-type reactor containing 0.6 m2 of photosensitizing material saturated with tris(4,7-diphenyl-1,10-phenanthroline)ruthenium(II) (ca. 2.0 g m(-2)). The optimum rheological conditions require laminar-to-transitional water flow in both prototypes. The fin-type system showed better inactivation efficiency than the coaxial reactor due to a more important photolytic contribution. The durability of the sensitizing materials was tested and the operational lifetime of the photocatalyst is at least three months without any reduction in the bacteria inactivation efficiency. Solar water disinfection with 1O2-generating films is demonstrated to be an effective technique for use in isolated regions of developing countries with high yearly average sunshine.

  6. Modeling relations between the composition and properties of French light water reactor waste containment glass

    Energy Technology Data Exchange (ETDEWEB)

    Ghaleb, D.; Dussossoy, J.L.; Fillet, C.; Pacaud, F.; Jacquet-Francillon, N.

    1994-12-31

    Models have been developed to calculate the density, molten-state viscosity and initial corrosion rate according to the chemical composition of glass formulations used to vitrify high-level fission product solutions from reprocessed light water reactor fuel. Developed from other published work, these models have been adapted to allow for the effects of platinoid (Ru, Pd, Rh) inclusions on the molten glass rheology. (authors). 15 refs., 10 figs., 1 tab.

  7. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    Energy Technology Data Exchange (ETDEWEB)

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  8. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

    Energy Technology Data Exchange (ETDEWEB)

    Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  9. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor.

    Science.gov (United States)

    Essadki, A H; Gourich, B; Vial, Ch; Delmas, H; Bennajah, M

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12mA/cm(2), but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H(2) microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  10. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  11. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  12. Analysis of transients in advanced heavy water reactor using lumped parameter models

    Energy Technology Data Exchange (ETDEWEB)

    Manmohan Pandey; Venkata Ramana Eaga; Sankar Sastry, P. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India); Gupta, S.K.; Lele, H.G.; Chatterjee, B. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    2005-07-01

    Full text of publication follows: Analysis of transients occurring in nuclear power plants, arising from the complex interplay between core neutronics and thermal-hydraulics, is important for their operation and safety. Numerical simulations of such transients can be carried out extensively at very low computational cost by using lumped parameter mathematical models. The Advanced Heavy Water Reactor (AHWR), being developed in India, is a vertical pressure tube type reactor cooled by boiling light water under natural circulation, using thorium as fuel and heavy water as moderator. In the present work, nonlinear and linear lumped parameter dynamic models for AHWR have been developed and validated with a distributed parameter model. The nonlinear lumped model is based on point reactor kinetics equations and one-dimensional homogeneous equilibrium model of two-phase flow. The distributed model is built with RELAP5/MOD3.2 code. Various types of transients have been simulated numerically, using the lumped model as well as RELAP5. The results have been compared and parameters tuned to make the lumped model match the distributed model (RELAP5) in terms of steady state as well as dynamic behaviour. The linear model has been derived by linearizing the nonlinear model for small perturbations about the steady state. Numerical simulations of transients using the linear model have been compared with results obtained from the nonlinear model. Thus, the range of validity of the linear model has been determined. Stability characteristics of AHWR have been investigated using the lumped parameter models. (authors)

  13. Nuclear reactor composite fuel assembly. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, D.M.; Cappiello, M.W.; Marr, D.R.; Omberg, R.P.

    1980-11-25

    A core and composite fuel assembly are described for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  14. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  15. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment.

  16. Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Junli Gou

    2009-01-01

    Full Text Available A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS, which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS, the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

  17. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  18. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor

    Energy Technology Data Exchange (ETDEWEB)

    Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: vaveri@kist.re.kr; Kim, Jae-Duck [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: jdkim@kist.re.kr; Lee, Jong-Chol [Agency for Defense Development (ADD), P.O. Box 35-1, Yuseong-gu, Daejeon (Korea, Republic of)]. E-mail: jcleeadd@hanafos.com

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG (HOC{sub 2}H{sub 4}){sub 2}S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  19. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor.

    Science.gov (United States)

    Veriansyah, Bambang; Kim, Jae-Duck; Lee, Jong-Chol

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG, (HOC(2)H(4))2S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  20. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  1. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  2. Solar radiation disinfection of drinking water at temperate latitudes: inactivation rates for an optimised reactor configuration.

    Science.gov (United States)

    Davies, C M; Roser, D J; Feitz, A J; Ashbolt, N J

    2009-02-01

    Solar radiation-driven inactivation of bacteria, virus and protozoan pathogen models was quantified in simulated drinking water at a temperate latitude (34 degrees S). The water was seeded with Enterococcus faecalis, Clostridium sporogenes spores, and P22 bacteriophage, each at ca 1x10(5) mL(-1), and exposed to natural sunlight in 30-L reaction vessels. Water temperature ranged from 17 to 39 degrees C during the experiments lasting up to 6h. Dark controls showed little inactivation and so it was concluded that the inactivation observed was primarily driven by non-thermal processes. The optimised reactor design achieved S90 values (cumulative exposure required for 90% reduction) for the test microorganisms in the range 0.63-1.82 MJ m(-2) of Global Solar Exposure (GSX) without the need for TiO2 as a catalyst. High turbidity (840-920 NTU) only reduced the S(90) value by 0.05). However, inactivation was significantly reduced for E. faecalis and P22 when the transmittance of UV wavelengths was attenuated by water with high colour (140 PtCo units) or a suboptimally transparent reactor lid (prob.waters and microorganisms. Although temperatures required for SODIS type pasteurization were not produced, non-thermal inactivation alone appeared to offer a viable means for reliably disinfecting low colour source waters by greater than 4 orders of magnitude on sunny days at 34 degrees S latitude.

  3. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  4. Flow injection spectrophotometry coupled with a crushed barium sulfate reactor column for the determination of sulfate ion in water samples

    OpenAIRE

    Burakham,Rodjana; Higuchi, Keiro; Oshima, Mitsuko; Grudpan, Kate; Motomizu, Shoji

    2004-01-01

    A new type of a reactor column, a crushed BaSO4 reactor column used for the flow injection spectrophotometric determination of sulfate ion using the exchange reaction of sulfate ion and barium-dimethylsulfonazo III is proposed. The column is very simple and economical. It can be continuously used for 8 h before washing with water for repeated usage of at least 1 month. The procedure is sensitive. Application to various water samples was demonstrated.

  5. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

  6. Study on mixed convective flow penetration into subassembly from reactor hot plenum in FBRs

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, J.; Ohshima, H.; Kamide, H.; Ieda, Y. [Power Reactor and Nuclear Fuel Development Corporation, Ibaraki (Japan)

    1995-09-01

    Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor`s upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis.

  7. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady

    2013-06-01

    Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

  8. Appendix to HDC 2118 design criteria 100-X reactor water plant, general description - section II

    Energy Technology Data Exchange (ETDEWEB)

    1952-03-29

    The factors responsible for the advances of 100-X compared with the older areas are: Simplification of the process, such as elimination of separate process water clearwells, by having the filtered water reservoirs perform that function. Combination of separate buildings into one building, such as combining filter pump house and process pump house. Use of electric standby. Use of higher capacity pumps and filter basins, and so fewer number of units. Centralization of control and operation. More compact arrangement of plant components. Use of waste heat for space heating, recovered from reactor effluent, backed up by steam plant.

  9. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  10. Membrane contactor/separator for an advanced ozone membrane reactor for treatment of recalcitrant organic pollutants in water

    Energy Technology Data Exchange (ETDEWEB)

    Chan, Wai Kit, E-mail: kekyeung@ust.hk [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Joueet, Justine; Heng, Samuel; Yeung, King Lun [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Schrotter, Jean-Christophe [Water Research Center of Veolia, Anjou Recherche, Chemin de la Digue, BP 76. 78603, Maisons Laffitte, Cedex (France)

    2012-05-15

    An advanced ozone membrane reactor that synergistically combines membrane distributor for ozone gas, membrane contactor for pollutant adsorption and reaction, and membrane separator for clean water production is described. The membrane reactor represents an order of magnitude improvement over traditional semibatch reactor design and is capable of complete conversion of recalcitrant endocrine disrupting compounds (EDCs) in water at less than three minutes residence time. Coating the membrane contactor with alumina and hydrotalcite (Mg/Al=3) adsorbs and traps the organics in the reaction zone resulting in 30% increase of total organic carbon (TOC) removal. Large surface area coating that diffuses surface charges from adsorbed polar organic molecules is preferred as it reduces membrane polarization that is detrimental to separation. - Graphical abstract: Advanced ozone membrane reactor synergistically combines membrane distributor for ozone, membrane contactor for sorption and reaction and membrane separator for clean water production to achieve an order of magnitude enhancement in treatment performance compared to traditional ozone reactor. Highlights: Black-Right-Pointing-Pointer Novel reactor using membranes for ozone distributor, reaction contactor and water separator. Black-Right-Pointing-Pointer Designed to achieve an order of magnitude enhancement over traditional reactor. Black-Right-Pointing-Pointer Al{sub 2}O{sub 3} and hydrotalcite coatings capture and trap pollutants giving additional 30% TOC removal. Black-Right-Pointing-Pointer High surface area coating prevents polarization and improves membrane separation and life.

  11. A modified firefly algorithm applied to the nuclear reload problem of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali Santos de; Schirru, Roberto, E-mail: ioliveira@con.ufrj.b, E-mail: schirru@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    The Nuclear Reactor Reload Problem (NRRP) is an issue of great importance and concern in nuclear engineering. It is the problem related with the periodic operation of replacing part of the fuel of a nuclear reactor. Traditionally, this procedure occurs after a period of operation called a cycle, or whenever the nuclear power plant is unable to continue operating at its nominal power. Studied for more than 40 years, the NRRP still remains a challenge for many optimization techniques due to its multiple objectives concerning economics, safety and reactor physics calculations. Characteristics such as non-linearity, multimodality and high dimensionality also make the NRRP a very complex optimization problem. In broad terms, it aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. The primary goal is to design fuel loading patterns (LPs) so that the core produces the required energy output in an economical way, without violating safety limits. Since multiple feasible solutions can be obtained to this problem, judicious optimization is required in order to identify the most economical among them. In this sense, this paper presents a new contribution in this area and introduces a modified firefly algorithm (FA) to perform LPs optimization for a pressurized water reactor. Based on the original FA introduced by Xin-She Yang in 2008, the proposed methodology seems to be very promising as an optimizer to the NRRP. The experiments performed and the comparisons with some well known best performing algorithms from the literature, confirm this statement. (author)

  12. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Andrey [Paul Scherrer Institut, Villigen (Switzerland); Degueldre, Claude, E-mail: claude.degueldre@psi.ch [Paul Scherrer Institut, Villigen (Switzerland); Kaufmann, Wilfried [Kernkraftwerk Leibstadt, Leibstadt (Switzerland)

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  13. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)

    2008-07-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  14. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  15. Degradation and detoxification of tar water from a gasification plant in a biogas reactor

    Energy Technology Data Exchange (ETDEWEB)

    Angelidaki, I.; Torry-Smith, M.; Petersen, L.; Ahring, B.K. [Denmarks Technical Univ., Dept. of Environmental Science and Engineering, Lyngby (Denmark)

    1998-08-01

    During gasification of biomass, toxic tar and tar-water (TW) are produced. The produced TW is saturated by soluble phenolic compound. In this project we have attempted detoxification of the produced TW by either wet oxidation where the TW is heated under pressure and addition of excess oxygen, and then subsequent treatment of the formed wet oxidized product in a biogas reactor. Alternatively, we treated the TW directly in a biogas reactor. Reduction in phenolic compounds, which constitute the major toxic compounds in TW, was used as a success parameter evaluating the detoxification of TW. The TW could successfully be degraded in a biogas reactor when co-digested with manure at a concentration up to 5%. Wet oxidized tar water (WOTW) could be degraded when added at a concentration of 30%. A biogas potential of approx. 0.19 L CH/g-VS was achieved for both the TW and WOTW. The biogas production per kg waste was 30 and 10 L CH/kg waste for the TW and the WOTW, respectively. The reason for the much lower methane production of the WOTW was that a large part of the organic content of the TW was converted to carbon dioxide during the wet oxidation process. The effluent concentrations of phenolic were in all cases much lower than the concentrations found in normally in undigested manure, showing that biomethanation of toxic wastes in co-digestion with manure could be an easy and cheap way to detoxify specific toxic wastes. (au)

  16. Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

    2017-05-15

    Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

  17. Storage of LWR (light-water-reactor) spent fuel in air

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, L.E.; Charlot, L.A.; Coleman, J.E. (Pacific Northwest Lab., Richland, WA (USA)); Knoll, R.W. (Johnson Controls, Inc., Madison, WI (USA))

    1989-12-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

  18. Solar disinfection of contaminated water: a comparison of three small-scale continuous flow reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gill, L. W.; McLoughlin, O. A.; McGuigan, K. G.; Duffy, E. F.; Kehole, S. C.; Al Touati, F.; Gernjak, W.; Oller, I.; Fernandez, P.; Malato, S.

    2004-07-01

    This paper compaes three different collector shapes for the disinfection of water heavily contaminated with E. coli (k-12). Tests were carried out in real sunlight using laboratory scale reactors to determine the performance of different reflectors of compound parabolic, parabolic and V-groove profile. Results have shown successful inactivation of E. coli from an initial concentration of 1x10''6 CFU/ml in one litre within 60 minutes in all three cases. Tests were also carried out to assess the improvement to disinfection which could be achieved using TiO2 coated Pyrex rods fixed within the reactors. This technique, however, only yielded a slight benefit to overall disinfection performance. (Author) 13 refs.

  19. Transactions of the Twenty-First Water Reactor Safety Information Meeting

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1993-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear indus