Start-up test of the prototype heavy water reactor 'FUGEN', (1)

International Nuclear Information System (INIS)

Ando, Hideki; Kawahara, Toshio

1982-01-01

The advanced thermal prototype reactor ''Fugen'' is a heavy water-moderated, boiling light water-cooled power reactor with electric output of 165 MW, which has been developed since 1966 as a national project. The start-up test was begun in March, 1978, being scheduled for about one year, and in March, 1979, it passed the final pre-use inspection and began the full scale operation. In this paper, the result of the start-up test of Fugen is reported. From the experience of the start-up test of Fugen, the following matters are important for the execution of start-up test. 1) Exact testing plan and work schedule, 2) the organization to perform the test, 3) the rapid evaluation of test results and the reflection to next testing plan, and 4) the reflection of test results to rated operation, regular inspection and so on. In the testing plan, the core characteristics peculiar to Fugen, and the features of heavy water-helium system, control system and other equipment were added to the contents of the start-up test of BWRs. The items of the start-up test were reactor physics test, plant equipment performance test, plant dynamic characteristic test, chemical and radiation measurement, and combined test. The organization to perform the start-up test, and the progress and the results of the test are reported. (Kako, I.)

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

International Nuclear Information System (INIS)

Manimaran, M.; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.

2015-01-01

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

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.

Conceptual design of reactor assembly of prototype fast breeder reactor

International Nuclear Information System (INIS)

Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

1996-01-01

The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

Prototype development and demonstration for response, emergency staging, communications, uniform management, and evacuation (R.E.S.C.U.M.E.) : R.E.S.C.U.M.E. prototype system design document.

Science.gov (United States)

2014-04-01

This report documents the System Design Document (SDD) for the prototype development and demonstration of the : Response, Emergency Staging, Communications, Uniform Management, and Evacuation (R.E.S.C.U.M.E.) application : bundle, with a focus on the...

Safety requirements expected to the prototype fast breeder reactor 'Monju'

International Nuclear Information System (INIS)

2014-11-01

In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

Monitoring the Thermal Power of Nuclear Reactors with a Prototype Cubic Meter Antineutrino Detector

Energy Technology Data Exchange (ETDEWEB)

Bernstein, A; Bowden, N; Misner, A; Palmer, T

2007-06-27

In this paper, we estimate how quickly and how precisely a reactor's operational status and thermal power can be monitored over hour to month time scales, using the antineutrino rate as measured by a cubic meter scale detector. Our results are obtained from a detector we have deployed and operated at 25 meter standoff from a reactor core. This prototype can detect a prompt reactor shutdown within five hours, and monitor relative thermal power to 3.5% within 7 days. Monitoring of short-term power changes in this way may be useful in the context of International Atomic Energy Agency's (IAEA) Reactor Safeguards Regime, or other cooperative monitoring regimes.

Development of a fresh plutonium fuel container for a prototype fast breeder reactor

International Nuclear Information System (INIS)

Ohtake, T.; Takahashi, S.; Mishima, T.; Kurakami, J.; Yamamoto, Y.; Ohuchi, Y.

1989-01-01

Japan gives a good deal of encouragement to development of a fast breeder reactor (which is considered as the most likely candidate for nuclear power generation) to secure long-term energy source. And, following an experimental fast breeder reactor Joyo, a prototype fast breeder reactor Monju is now under vigorous construction. Related to development of the prototype fast breeder reactor, it is necessary and important to develop transport container which is used for transporting fresh fuel assemblies from Plutonium Fuel Production Facility to the Monju power plant. Therefore, the container is now being developed by Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, shipment and vibration tests, handling performance tests, shielding performance tests and prototype container tests are executed with prototype containers fabricated according to a final design, in order to experimentally confirm soundness of transport container and its contents, and propriety of design technique. This paper describes the summary of general specifications and structures of this container and the summary of preliminary safety analysis of package

Test reactor: basic to U.S. breeder reactor development

International Nuclear Information System (INIS)

Miller, B.J.; Harness, A.J.

1975-01-01

Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

International Nuclear Information System (INIS)

Conlin, J.A.

1985-08-01

A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

International Nuclear Information System (INIS)

Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

1999-09-01

A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

Challenges and achievements - Prototype Fast Breeder Reactor construction

International Nuclear Information System (INIS)

Subramani, V.A.; Dhere, S.S.; Manoharan, V.; Subbaraman, P.

2010-01-01

Prototype fast breeder reactor presently under construction poses several challenges in materials, design and construction. The civil structure and equipment are of very large size and complex in nature. This paper presents the features of the design and construction of the PFBR excavation, raft, civil structure of the nuclear island connected buildings and reactor vault. This paper also brings out the details of the large size equipment of special stainless steel and handling structure for their lifting and placement inside the reactor vault. The paper is divided into three parts viz. introduction, challenges and achievements during construction of civil structures and erection of large size components. (author)

Efficient infectious cell culture systems of the hepatitis C virus (HCV) prototype strains HCV-1 and H77.

Science.gov (United States)

Li, Yi-Ping; Ramirez, Santseharay; Mikkelsen, Lotte; Bukh, Jens

2015-01-01

The first discovered and sequenced hepatitis C virus (HCV) genome and the first in vivo infectious HCV clones originated from the HCV prototype strains HCV-1 and H77, respectively, both widely used in research of this important human pathogen. In the present study, we developed efficient infectious cell culture systems for these genotype 1a strains by using the HCV-1/SF9_A and H77C in vivo infectious clones. We initially adapted a genome with the HCV-1 5'UTR-NS5A (where UTR stands for untranslated region) and the JFH1 NS5B-3'UTR (5-5A recombinant), including the genotype 2a-derived mutations F1464L/A1672S/D2979G (LSG), to grow efficiently in Huh7.5 cells, thus identifying the E2 mutation S399F. The combination of LSG/S399F and reported TNcc(1a)-adaptive mutations A1226G/Q1773H/N1927T/Y2981F/F2994S promoted adaptation of the full-length HCV-1 clone. An HCV-1 recombinant with 17 mutations (HCV1cc) replicated efficiently in Huh7.5 cells and produced supernatant infectivity titers of 10(4.0) focus-forming units (FFU)/ml. Eight of these mutations were identified from passaged HCV-1 viruses, and the A970T/I1312V/C2419R/A2919T mutations were essential for infectious particle production. Using CD81-deficient Huh7 cells, we further demonstrated the importance of A970T/I1312V/A2919T or A970T/C2419R/A2919T for virus assembly and that the I1312V/C2419R combination played a major role in virus release. Using a similar approach, we found that NS5B mutation F2994R, identified here from culture-adapted full-length TN viruses and a common NS3 helicase mutation (S1368P) derived from viable H77C and HCV-1 5-5A recombinants, initiated replication and culture adaptation of H77C containing LSG and TNcc(1a)-adaptive mutations. An H77C recombinant harboring 19 mutations (H77Ccc) replicated and spread efficiently after transfection and subsequent infection of naive Huh7.5 cells, reaching titers of 10(3.5) and 10(4.4) FFU/ml, respectively. Hepatitis C virus (HCV) was discovered in 1989 with

R&D on high-power dc reactor prototype for ITER poloidal field converter

Energy Technology Data Exchange (ETDEWEB)

Li, Chuan [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Song, Zhiquan; Fu, Peng [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Zhang, Ming, E-mail: zhangming@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yu, Kexun [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Qin, Xiuqi [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China)

2015-10-15

Highlights: • A new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented. • Theoretical analysis, finite-element simulation and prototype test verification are applied on the design. • The results of temperature rise and transient fault current test of prototypes are introduced and analyzed. • The success of tests demonstrates that the proposed structure is of high reliability and availability. - Abstract: This paper mainly introduces the research and development (R&D) of the high-power dc reactor prototype, whose functions are to limit the circulating current and ripple current in the ITER poloidal field (PF) converter. It needs to operate at rated large direct current 27.5 kA and withstand peak fault current up to 175 kA. Therefore, in order to meet the special requirements of the dynamic and thermal stability, a new prototype design structure of dry-type air-core water-cooling reactor with epoxy resin casting technique is presented, which is based on the theoretical analysis, finite-element simulation calculation and small prototype test verification. Now the full prototype has been fabricated by China industry, and the dynamic and thermal stability tests of the prototype have also been accomplished successfully. The test results are in compliance with the design and it shows the availability and feasibility of the proposed design, which may be a reference for relevant applications.

Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

International Nuclear Information System (INIS)

Ishiyama, Shintaro; Muto, Yasushi; Kato, Yasuyoshi; Nishio, Satoshi; Hayashi, Takumi; Nomoto, Yasunobu

2008-01-01

Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO 2 (S-CO 2 ) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480degC, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO 2 turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO 2 cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m 3 and 7240 m 3 for the steam turbine system and S-CO 2 turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO 2 than in H 2 O. Therefore, the S-CO 2 turbine system is recommended to the fusion reactor system than the steam turbine system. (author)

Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

International Nuclear Information System (INIS)

Perry, E.F.

1995-01-01

The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

Total decay heat estimates in a proto-type fast reactor

International Nuclear Information System (INIS)

Sridharan, M.S.

2003-01-01

Full text: In this paper, total decay heat values generated in a proto-type fast reactor are estimated. These values are compared with those of certain fast reactors. Simple analytical fits are also obtained for these values which can serve as a handy and convenient tool in engineering design studies. These decay heat values taken as their ratio to the nominal operating power are, in general, applicable to any typical plutonium based fast reactor and are useful inputs to the design of decay-heat removal systems

Design of micro-reactors and solar photocatalytic prototypes

International Nuclear Information System (INIS)

Flores E, R.M.; Hernandez H, M.; Perusquia del Cueto, M.R.; Bonifacio M, J.; Jimenez B, J.; Ortiz O, H.B.; Castaneda J, G.; Lugo H, M.

2007-01-01

In the ININ is carried out research in heterogeneous photocatalysis using artificial light for to degrade organic compounds. In this context, it is sought to use the solar radiation as energy source to knock down costs. Of equal form it requires to link the basic and applied research. For it, a methodology that allows to design and to build micro-reactors and plants pilot has been developed, like previous step, to request external supports and to a future commercialization. The beginning of these works gave place to the partial construction of a prototype of photocatalytic reactor of the cylinder-parabolic composed type (CPC)

Safety demonstration test (SR-3/S1C-3/S2C-3/SF-2) plan using the HTTR. Contract research

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji; Tochio, Daisuke; Ohwada, Hiroyuki

2005-03-01

Safety demonstration tests using the HTTR are to be conducted from the FY2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR that is one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3, S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was verified that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram. (author)

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

2003-01-01

Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

Energy Technology Data Exchange (ETDEWEB)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

A new physics design of control safety rods for prototype fast breeder reactor

International Nuclear Information System (INIS)

Devan, K.; Riyas, A.; Alagan, M.; Mohanakrishnan, P.

2008-01-01

The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B 4 C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B 4 C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B 4 C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B 4 C sections in top and bottom. There is significant savings in the initial inventory of enriched B 4 C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles

Irradiation of an uranium silicide prototype in RA-3 reactor

International Nuclear Information System (INIS)

Calabrese, R.; Estrik, G.; Notari, C.

1996-01-01

The factibility of irradiation of an uranium silicide (U 3 Si 2 ) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U 3 O 8 as fissible material. The enrichment of both standard and prototype is the same: 20% U 235 and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs

Some post operational adjustments to the prototype fast reactor at Dounreay

International Nuclear Information System (INIS)

Lunt, A.R.W.

1979-01-01

Prior to and during the initial operation of the Prototype Fast Reactor at Dounreay certain features have been considered to be in need of adjustment to provide better operating characteristics. This article describes the work done to support the consequential changes of operational techniques and plant design in the following areas: maintenance of dry conditions at the superheater steam inlets, the temperature control of the reactor roof, and the introduction of a system enabling the reactor to continue running after a turbine trip. (author)

Description of the prototype fast reactor at Dounreay

International Nuclear Information System (INIS)

Jensen, S.E.; Oelgaard, P.L.

1995-12-01

The Prototype Fast Reactor (PFR) at Dounreay, UK, started operation in 1975 and was closed down in 1994. The present report contains a description of the PFR nuclear power plant, based on information available in literature and on information supplied during a visit to the plant. The report covers a description of the site and plant arrangement, the buildings and structures, the reactor core and other vessel internals, the control system, the main cooling system, the decay heat removal system, the emergency core cooling system, the containment system, the steam and power conversion system, the fuel handling system, plant safety features, the control and instrumentation systems and the sodium purification systems. (au) 16 refs

Nuclear instrumentation systems in prototype fast breeder reactor

International Nuclear Information System (INIS)

Vijayakumaran, P.M.; Nagaraj, C.P.; Paramasivan-Pillai, C.; Ramakrishnan, R.; Sivaramakrishna, M.

2004-01-01

The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection and Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM (safety action) to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident. (authors)

Sodium tests on an integrated purification prototype for a sodium-cooled fast breeder reactor

International Nuclear Information System (INIS)

Abramson, R.

1984-04-01

This paper describes sodium tests performed on the integrated primary sodium purification prototype of the Creys Malville Super Phenix 1 fast breeder reactor. These tests comprised: - hydrostatic test, - thermal tests, - handling tests. They enabled a number of new technological arrangements to be qualified and provided the necessary information for the design and construction of the Super Phenix 1 purification units

Fault tolerant distributed real time computer systems for I and C 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.

2014-03-15

Highlights: • Architecture of distributed real time computer system (DRTCS) used in I and C of PFBR is explained. • Fault tolerant (hot standby) architecture, fault detection and switch over are detailed. • Scaled down model was used to study functional and performance requirements of DRTCS. • Quality of service parameters for scaled down model was critically studied. - Abstract: Prototype fast breeder reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Three-tier architecture is adopted for instrumentation and control (I and C) of PFBR wherein bottom tier consists of real time computer (RTC) systems, middle tier consists of process computers and top tier constitutes of display stations. These RTC systems are geographically distributed and networked together with process computers and display stations. Hot standby architecture comprising of dual redundant RTC systems with switch over logic system is deployed in order to achieve fault tolerance. Fault tolerant dual redundant network connectivity is provided in each RTC system and TCP/IP protocol is selected for network communication. In order to assess the performance of distributed RTC systems, scaled down model was developed with 9 representative systems and nearly 15% of I and C signals of PFBR were connected and monitored. Functional and performance testing were carried out for each RTC system and the fault tolerant characteristics were studied by creating various faults into the system and observed the performance. Various quality of service parameters like connection establishment delay, priority parameter, transit delay, throughput, residual error ratio, etc., are critically studied for the network.

A Takagi–Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

Directory of Open Access Journals (Sweden)

Yue Yuan

2017-08-01

Full Text Available Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

A Takagi-Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

Energy Technology Data Exchange (ETDEWEB)

Yuan, Yue [Institute of Nuclear and New Energy Technology, Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing (China); Coble, Jamie [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

2017-08-15

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S) fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

Evolution of actinides in ThO2 blanket of prototype fast breeder reactor

International Nuclear Information System (INIS)

Bachchan, Abhitab; Riyas, A.; Devan, K.; Puthiyavinayagam, P.

2015-01-01

The third stage of India's nuclear program focuses on fissile fuel production through Th- 233 U cycle in view of the better abundance and relative merits of thorium. For early introduction of Thorium into the nuclear energy system, several R and D program has started to find the best possible route of thorium utilization. Towards this, efforts were made to assess the feasibility of Th-U cycle in a fast spectrum reactor like Prototype Fast Breeder Reactor (PFBR). The effect on core neutronic parameters and actinide evolution with the replacement of depleted UO 2 in the PFBR blanket SA with thorium oxide has been studied using 3-D diffusion code FARCOB. Study shows that by the introduction of thorium blanket, core excess reactivity is coming down by ∼ 535 pcm and core breeding ratio is slightly lower than conventional oxide blanket. The distribution of region wise power production is slightly changed. Power from radial blanket is reduced from 3% to 2% while the core-1 power is increased from 49 % to 50 %. The estimated 233 U production is 7.6, 11.5 and 14.1 kg/t with 180, 360 and 540 days of irradiation respectively. (author)

Operability design review of prototype large breeder reactor (PLBR) designs. Final report, September 1981

International Nuclear Information System (INIS)

Beakes, J.H.; Ehman, J.R.; Jones, H.M.; Kinne, B.V.T.; Price, C.M.; Shores, S.P.; Welch, J.K.

1981-09-01

Prototype Large Breeder Reactor (PLBR) designs were reviewed by personnel with extensive power plant operations experience. Fourteen normal and off-normal events, such as startup, shutdown, refueling, reactor scram and loss of feedwater, were evaluated using an operational evaluation methodology which is designed to facilitate talk-through sessions on operational events. Human factors engineers participated in the review and assisted in developing and refining the review methodologies. Operating experience at breeder reactor facilities such as Experimental Breeder Reactor-II (EBR-II), Enrico Fermi Atomic Power Plant - Unit 1, and the Fast Flux Test Facility (FFTF) was gathered, analyzed, and used to determine whether lessons learned from operational experience had been incorporated into the PLBR designs. This eighteen month effort resulted in approximately one hundred specific recommendations for improving the operability of PLBR designs

Concept of data storage prototype for Super-C-Tau factory detector

International Nuclear Information System (INIS)

Maximov, D.A.

2017-01-01

The physics program of experiments at the Super- c τ factory with a peak luminosity of 10 35 cm −2 s −1 leads to high requrements for Data Acquisition and Data Storage systems. Detector data storage is one of the key component of the detector infrastructure, so it must be reliable, highly available and fault tolerant shared storage. It is mostly oriented (from end user point of view) for sequential but mixed read and write operations and is planed to store large data blocks (files). According to CDR of Super-C-Tau factory detector data storage must have very high performance (up to 1 Tbps in both directions simultaneously) and have significant volume (tens and hundreds of Petabytes). It is decided to build a series of prototypes with growing capabilities to investigate storage and neighboring technologies. First prototype of data storage is aimed to develop and test basic components of detector data storage system such as storage devices, networks and software. This prototype is designed to be capable to work with data rate of order 10 Gbps. It is estimated that about 5 modern computers with about 50 disks in total should be enough to archive required performance. The prototype will be based on Ceph storage technology. Ceph is a distributed storage system which allows to create storage solutions with very flexible design, high availability and scalability.

Operation management of the prototype heavy water reactor 'Fugen'

International Nuclear Information System (INIS)

Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi

1983-01-01

The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported. (Kako, I.)

Reactivity monitoring for safety purposes on the UK prototype fast reactor

International Nuclear Information System (INIS)

Lord, D.J.; Wilkes, D.J.

1987-01-01

The small size and high rating of the liquid metal cooled fast breeder reactor (LMFBR) make the provision of safety related instrumentation for individual subassemblies both difficult and expensive. Global monitoring of the core is thus very attractive. Reactivity monitoring is an important part of such global monitoring. Reactivity monitoring on a short timescale (a few seconds) is used on the UK Prototype Fast Reactor (PFR) as a trip parameter and long-term reactivity monitoring is being developed as a means of providing early warning of slowly developing faults. Results are presented from PFR to demonstrate the capabilities of reactivity monitoring in an operational fast reactor power station. (author)

Upgraded prototype-reactor internal pump for ABWR

International Nuclear Information System (INIS)

Kumagai, Mikio; Amemori, Shiro; Saito, Takehiko

1988-01-01

In 1983, Toshiba, using their own technology, manufactured a commercial grade reactor internal pump (RIP). Recently, however, a licensing agreement with KSB of West Germany covering the RIP technology, has combined the know-how of KSB with Toshiba's technology to produce a truly high-quality prototype RIP. The pump produces the required coreflow for ABWR at low speed and with high efficiency, and simply by increasing the pump speed to the prior level, the coreflow can be further increased for such advantages as improved fuel cycle economy. Here, the advanced features and test results of the RIP are summarized. (author)

Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

2003-01-01

The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

ASN’s actions in GEN IV reactors and Sodium Fast Reactors (SFR)

International Nuclear Information System (INIS)

Belot, Clotilde

2013-01-01

The ASN is involved in 3 actions concerning GEN IV: • Overview of nuclear reactor GEN IV systems; • Specific analysis about transmutation; • Prototype reactor ASTRID (SFR). Furthermore theses actions are in the beginning (no conclusions or results available)

Comparative Studies of Core Thermal Hydraulic Design Methods for the Prototype Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Sun Rock; Lim, Jae Yong; Kim, Sang Ji

2013-01-01

In this work, various core thermal-hydraulic design methods, which have arisen during the development of a prototype SFR, are compared to establish a proper design procedure. Comparative studies have been performed to determine the appropriate design method for the prototype SFR. The results show that the minimization method show a lower cladding midwall temperature than the fixed outlet temperature methods and superior thermal safety margin with the same coolant flow. The Korea Atomic energy Research Institute (KAERI) has performed a conceptual SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damages in SFR subassemblies are arisen from a creep induced failure. The creep limit is evaluated based on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, the core thermalhydraulic design method, which eventually determines the cladding temperature, is highly important to assure a safe and reliable operation of the reactor systems

Multiple recycling of fuel in prototype fast breeder reactor in a closed ...

Indian Academy of Sciences (India)

Our previous study in this regard for the prototype fast breeder reactor ... This study aims at finding the feasibility of multiple recycling of PFBR fuel with external ...... maximum allowable Pu content in fuel based on chemistry/metallurgical ...

Development of prototype reactor maintenance. (2) Application to piping support of sodium-cooled reactor prototype

International Nuclear Information System (INIS)

Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji; Ito, Takaya; Yamaguchi, Akira

2017-01-01

A maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of piping supports could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports. (author)

Design study of prototype accelerator and MeV test facility for demonstration of 1 MeV, 1 A negative ion beam production

International Nuclear Information System (INIS)

Inoue, Takashi; Hanada, Masaya; Miyamoto, Kenji; Ohara, Yoshihiro; Okumura, Yoshikazu; Watanabe, Kazuhiro; Maeno, Shuichi.

1994-08-01

In fusion reactors such as ITER, a neutral beam injector of MeV class beam energy and several tens MW class power is required as one of candidates of heating and current drive systems. However, the beam energy of existing high power accelerators are one order of magnitude lower than the required value. In order to realize a neutral beam injector for the fusion reactor, 'Proof-of-Principle' of such high energy acceleration is a critical issue at a reactor relevant beam current and pulse length. An accelerator and an accelerator facility which are necessary to demonstrate the Proof-of-Principle acceleration of negative ion beams up to 1 MeV, have been designed in the present study. The accelerator is composed of a cesium-volume type ion source and a multi-stage electrostatic acceleration system [Prototype Accelerator]. A negative hydrogen ion beam with the current of about one ampere (1 A) can be accelerated up to 1 MeV at a low operating pressure. Two types of acceleration system, a multi-multi type and a multi-single type, have been studied. The test facility has sufficient capability for the test of the Prototype Accelerator [MeV Test Facility]. The dc high voltage generator for negative ion acceleration is a Cockcroft-Walton type and capable of delivering 1 A at 1 MV (=1 MW) for 60 s. High voltage components including Prototype Accelerator are installed in a SF 6 vessel pressurized at 6 kg/cm 2 to overcome high voltage gradients. The vessel and the beamline are installed in a X-ray shield. (author)

Efficient infectious cell culture systems of the hepatitis C virus (HCV) prototype strains HCV-1 and H77

DEFF Research Database (Denmark)

Li, Yi-Ping; Ramirez, Santseharay; Mikkelsen, Lotte

2015-01-01

UNLABELLED: The first discovered and sequenced hepatitis C virus (HCV) genome and the first in vivo infectious HCV clones originated from the HCV prototype strains HCV-1 and H77, respectively, both widely used in research of this important human pathogen. In the present study, we developed...... efficiently after transfection and subsequent infection of naive Huh7.5 cells, reaching titers of 10(3.5) and 10(4.4) FFU/ml, respectively. IMPORTANCE: Hepatitis C virus (HCV) was discovered in 1989 with the cloning of the prototype strain HCV-1 genome. In 1997, two molecular clones of H77, the other HCV...... prototype strain, were shown to be infectious in chimpanzees, but not in vitro. HCV research was hampered by a lack of infectious cell culture systems, which became available only in 2005 with the discovery of JFH1 (genotype 2a), a genome that could establish infection in Huh7.5 cells. Recently, we...

Flow-induced vibration test of an advanced water reactor model. Pt. 1. Turbulence-induced forcing function

International Nuclear Information System (INIS)

Au-Yang, M.K.; Brenneman, B.; Raj, D.

1995-01-01

A 1:9 scale model of a proposed advanced water reactor was tested for flow-induced vibration. The main objectives of this test were: (1) to derive an empirical equation for the turbulence forcing function which can be applied to the full-sized prototype; (2) to study the effect of viscosity on the turbulence; (3) to verify the ''superposition'' assumption widely used in dynamic analysis of weakly coupled fluid-shell systems; and (4) to measure the shell responses to verify methods and computer programs used in the flow-induced vibration analysis of the prototype. This paper describes objectives (1), (2), and (3); objective (4) will be discussed in a companion paper.The turbulence-induced fluctuating pressure was measured at 49 locations over the surface of a thick-walled, non-responsive scale model of the reactor vessel/core support cylinders. An empirical equation relating the fluctuating pressure, the frequency, and the distance from the inlet nozzle center line was derived to fit the test data. This equation involves only non-dimensional, fluid mechanical parameters that are postulated to represent the full-sized, geometrically similar prototype. While this postulate cannot be verified until similar measurements are taken on the full-sized unit, a similar approach using a 1:6 scale model of a commercial pressurized water reactor was verified in the mid-1970s by field measurements on the full-sized reactor. (orig.)

Design of micro-reactors and solar photocatalytic prototypes; Diseno de micro-reactores y prototipos fotocataliticos solares

Energy Technology Data Exchange (ETDEWEB)

Flores E, R.M.; Hernandez H, M.; Perusquia del Cueto, M.R.; Bonifacio M, J.; Jimenez B, J.; Ortiz O, H.B.; Castaneda J, G.; Lugo H, M. [ININ, Km. 36.5 Carr. Mexico-Toluca, 52750 La Marquesa, Ocoyoacac (Mexico)]. e-mail: rmfe@nuclear.inin.mx

2007-07-01

In the ININ is carried out research in heterogeneous photocatalysis using artificial light for to degrade organic compounds. In this context, it is sought to use the solar radiation as energy source to knock down costs. Of equal form it requires to link the basic and applied research. For it, a methodology that allows to design and to build micro-reactors and plants pilot has been developed, like previous step, to request external supports and to a future commercialization. The beginning of these works gave place to the partial construction of a prototype of photocatalytic reactor of the cylinder-parabolic composed type (CPC)

Prototype development and demonstration for response, emergency staging, communications, uniform management, and evacuation (R.E.S.C.U.M.E.) : R.E.S.C.U.M.E. final functional and performance requirements.

Science.gov (United States)

2014-01-01

This document provides the high-level functional and performance requirements for the Prototype Development and Demonstration : of a R.E.S.C.U.M.E. system. The requirements included in this document are based upon those that can be found in previous ...

Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

International Nuclear Information System (INIS)

1989-09-01

The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

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

Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

International Nuclear Information System (INIS)

Wade, D.C.

1990-01-01

Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

Generation IV reactors and the ASTRID prototype: lessons from the Fukushima accident

International Nuclear Information System (INIS)

Gauche, F.

2012-01-01

In France, the ASTRID prototype is an industrial demonstrator of a sodium-cooled fast neutron reactor (SFR), fulfilling the criteria for Generation IV reactors. ASTRID will meet safety requirements as stringent as for third generation reactors, and it takes into account lessons from the Fukushima accident. The objectives are to reinforce the robustness of the safety demonstration for all safety functions. ASTRID will feature an innovative core with a negative sodium void coefficient, it will take advantage of the large thermal inertia of SFR for decay heat removal, and will provide for a design either eliminating the sodium-water reaction, or guaranteeing no consequences for safety in case such reaction would take place. (author)

Risk-oriented analysis on the German prototype fast breeder reactor SNR-300

International Nuclear Information System (INIS)

Bayer, A.; Koeberlein, K.; Gesellschaft fuer Reaktorsicherheit, Garching, Germany)

1984-01-01

On request of a fact-finding committee of the German Federal Parliament, a risk-oriented analysis on the SNR-300, the German prototype fast breeder reactor, has been performed to allow a pragmatic safety comparison of the SNR-300 and a modern light-water reactor. Results of the technical plant analysis have been summarized in seven release categories. Accident consequences have been calculated for the actual site at Kalkar/Rhine. The results indicate that for the SNR-300 both the frequency of major accidents and the consequences of accidents are smaller than for the pressurized-water reactor analyzed in the German Risk Study. This article summarizes the methods and main results of the analysis of the SNR-300

Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

International Nuclear Information System (INIS)

Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

1976-01-01

The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

Complete cDNA sequence of human complement C1s and close physical linkage of the homologous genes C1s and C1r

International Nuclear Information System (INIS)

Tosi, M.; Duponchel, C.; Meo, T.; Julier, C.

1987-01-01

Overlapping molecular clones encoding the complement subcomponent C1s were isolated from a human liver cDNA library. The nucleotide sequence reconstructed from these clones spans about 85% of the length of the liver C1s messenger RNAs, which occur in three distinct size classes around 3 kilobases in length. Comparisons with the sequence of C1r, the other enzymatic subcomponent of C1, reveal 40% amino acid identity and conservation of all the cysteine residues. Beside the serine protease domain, the following sequence motifs, previously described in C1r, were also found in C1s: (a) two repeats of the type found in the Ba fragment of complement factor B and in several other complement but also noncomplement proteins, (b) a cysteine-rich segment homologous to the repeats of epidermal growth factor precursor, and (c) a duplicated segment found only in C1r and C1s. Differences in each of these structural motifs provide significant clues for the interpretation of the functional divergence of these interacting serine protease zymogens. Hybridizations of C1r and C1s probes to restriction endonuclease fragments of genomic DNA demonstrate close physical linkage of the corresponding genes. The implications of this finding are discussed with respect to the evolution of C1r and C1s after their origin by tandem gene duplication and to the previously observed combined hereditary deficiencies of Clr and Cls

Upgrade of VR-1 training reactor I and C

International Nuclear Information System (INIS)

Kropik, M.; Matejka, K.; Chab, V.

2003-01-01

The contribution describes the upgrade of the VR-1 training reactor I and C (Instrumentation and Control). The reactor was put into operation in the 1990, and its I and C seems to be obsolete now. The new I and C utilises the same digital technology as the old one. The upgrade has been done gradually during holidays in order not to disturb the reactor utilisation during teaching and training. The first stage consisted in the human-machine interface and the control room upgrade in 2001. A new operator's desk, displays, indicators and buttons were installed. Completely new software and communication interface to the present I and C were developed. During the second stage in 2002, new control rod drivers and safety circuits were installed. The rod motors were replaced and necessary mechanical changes on the control rod mechanism, induced by the utilisation of the new motor, were done. The new safety circuits utilise high quality relays with forced contacts to guarantee high reliability of their operation. The third stage, the control system upgrade is being carried out now. The new control system is based on an industrial PC mounted in a 19 inch crate. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. A large amount of work has been devoted to the software requirements to specify all dependencies, modes and permitted actions, safety measures, etc. The Department took an active part in the setting of software requirements and later in verification and validation of the software and the whole control system. Finally, a new protection system consisting of power measuring and power protection channels will be installed in 2004 or 2005. (author)

Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Keddar, A.

1982-08-01

The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the 240 Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies

Prototype fast reactor steam generator unit pressure vessel repairs

International Nuclear Information System (INIS)

Daniels, B.D.; Green, D.; Henderson, J.D.C.

1993-01-01

The prototype fast reactor at Dounreay has experienced a number of unscheduled shutdowns due to leaking reheater and superheater shell welds. There was a need to determine the cracking mechanism and to design a general repair technique simultaneously. Detailed investigations revealed that the crack locations correlated with the positions of rectification welds made at the time of vessel manufacture. A creep crack growth mechanism was identified; this requires through wall residual stress for through cracks to develop. A repair technique has been devised and successfully applied to the sites of a number of leaks. (author)

A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

Engineering and planning for reactor 105-C interim safe storage project subcontract no. 0100C-SC-G0001 conceptual design report. Volume 1

International Nuclear Information System (INIS)

1996-04-01

The 105-C Reactor, one of eight surplus production reactors at the Hanford Site, has been proposed by the U.S. Department of Energy, Richland, Operations Office to be the first large-scale technology demonstration project in the decontamination and decommissioning (D ampersand D) focus area as part of the project for dismantlement and interim safe storage. The 105-C Reactor will be placed in an interim safe storage condition, then undergo the decontamination and decommissioning phase. After D ampersand D, the reactor will be placed in long- term safe storage. This report provides the conceptual design for these activities

Application of PSA level 1 for the Fugen prototype ATR plant

International Nuclear Information System (INIS)

Ikeda, H.; Iguchi, Y.; Sotsu, M.; Seki, O.; Satou, S.

1997-01-01

This paper presents application of PSA Level-1 for Prototype Advanced Thermal Reactor Plant ''Fugen'' in consideration of the design characteristics of such reactor and verify the safety aspect of Fugen using thus established procedure, ATR resembles the boiling water reactor (BWR) in a number of points, but there are also some differences between the ATR and the BWR. Therefore. PSA procedure have been established by taking such difference into consideration and by referring to experience of PSA in USA and Japan. Moreover, the core damage frequency was calculated on Fugen by using thus established procedure. As a result, it was verified that results including the maximum value of the uncertainty estimation were found to be quite satisfactory against the target value of reactor damage frequency defined by the International Atomic Energy Agency (IAEA). (author)

Operation management of the prototype heavy water reactor 'Fugen'

Energy Technology Data Exchange (ETDEWEB)

Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

1983-09-01

The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported.

Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

Energy Technology Data Exchange (ETDEWEB)

Hess, J. D.; Gauld, I. C.; Gulliford, J.; Hill, I.; Okajima, S.

2017-01-01

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Project (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.

CCSDS SM and C Mission Operations Interoperability Prototype

Science.gov (United States)

Lucord, Steven A.

2010-01-01

This slide presentation reviews the prototype of the Spacecraft Monitor and Control (SM&C) Operations for interoperability among other space agencies. This particular prototype uses the German Space Agency (DLR) to test the ideas for interagency coordination.

A small research reactor for the 1980's

International Nuclear Information System (INIS)

Baglin, C.; Collis-Smith, J.A.; Mitchell, B.; Roskilly, T.

1978-01-01

In 1960, GEC together with Imperial College, designed and built the Consort reactor which is still in daily use at the London University Reactor Centre, Silwood Park. In 1977, GEC-REL chose the Consort reactor as a prototype for the development of a modern swimming pool research reactor, designed to meet the needs of countries or organisations starting in the field of Nuclear Technology. This paper outlines some of the topics which arose in the course of this project. (author)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Calibration of new I and C at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, Martin; Jurickova, Monika

2011-01-01

The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

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.

Annual technical report of the prototype fast breeder reactor Monju. 2013

International Nuclear Information System (INIS)

2014-08-01

The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the principal achievements and the data related to the plant management in Monju in fiscal 2013. From the aspect of the management and design evaluation, the following items are summarized: 1) Current status of coping with the order from NRA to alter the safety regulations. 2) Implementation status of reformation of Monju. 3) Current status of the additional geological surveys of crush zones at the Monju site. 4) Development of core seismic assessment method for FBR. Then, from the aspect of the operation and maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management (Part 2). 2) Performance confirmation of the failed fuel detection and location system. 3) Deviation from the limiting conditions for operation in the emergency diesel generator periodic test and so on. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

Energy Technology Data Exchange (ETDEWEB)

Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G. [N.A. Dollezhal Institute ' NIKIET' , PO Box 788, Moscow, 101000 (Russian Federation)

2006-07-01

This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with {beta}{sub eff}, low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

International Nuclear Information System (INIS)

Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G.

2006-01-01

This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with β eff , low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

Project management plan for the 105-C Reactor interim safe storage project. Revision 1

International Nuclear Information System (INIS)

Miller, R.L.

1997-01-01

In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

Demonstrated operational and inherent safety of the prototype fast reactor (PFR)

International Nuclear Information System (INIS)

Smedley, J.A.; Gregory, C.V.; Judd, A.M.

1983-01-01

The Prototype Fast Reactor (PFR) is sited at Dounreay, on the north coast of Scotland in the United Kingdom, and has been in operation since 1974. Three aspects of the safety of the reactor are described, including the all-important practical consideration of operational safety, a demonstration of the limited consequences of a sodium/water reaction in a steam generator and the ability of the reactor to protect itself against highly improbable incidents. Attention is drawn to the low radiation levels in the plant and the correspondingly low dose rate to personnel. A feature of PFR operation has been the stable and predictable behaviour of its core together with the high degree of reliability exhibited by the engineered safety system. No failures have occurred within the standard driver charge but two experimental fuel pins suffered cladding failure, which was detected easily by the fission gas and delayed neutron detection systems. In the steam generating units sodium and water are separated by the single steel wall of the steam tubes. Although no under-sodium leak has occurred, an experimental programme is continuing and demonstrates that were any such leak to occur its consequences would be containable and would not result in the release of sodium to the environment or any breach of the reactor containment. The final section describes the inherent safety features of the reactor which enable it to survive a range of very improbable incidents even when the engineered safeguards fail. The features considered are natural circulation, which has been demonstrated by reactor experiment; the reactor's negative power coefficient, which, for example, enables the reactor to survive a complete loss of heat sink; and the durability of the fuel pins, demonstrated by a series of boiling experiments in the Dounreay Fast Reactor (DFR). (author)

DOE's annealing prototype demonstration projects

International Nuclear Information System (INIS)

Warren, J.; Nakos, J.; Rochau, G.

1997-01-01

One of the challenges U.S. utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels (RPVs). As a nuclear plant operates, its RPV is exposed to neutrons. For certain plants, this neutron exposure can cause embrittlement of some of the RPV welds which can shorten the useful life of the RPV. This RPV embrittlement issue has the potential to affect the continued operation of a number of operating U.S. pressurized water reactor (PWR) plants. However, RPV material properties affected by long-term irradiation are recoverable through a thermal annealing treatment of the RPV. Although a dozen Russian-designed RPVs and several U.S. military vessels have been successfully annealed, U.S. utilities have stated that a successful annealing demonstration of a U.S. RPV is a prerequisite for annealing a licensed U.S. nuclear power plant. In May 1995, the Department of Energy's Sandia National Laboratories awarded two cost-shared contracts to evaluate the feasibility of annealing U.S. licensed plants by conducting an anneal of an installed RPV using two different heating technologies. The contracts were awarded to the American Society of Mechanical Engineers (ASME) Center for Research and Technology Development (CRTD) and MPR Associates (MPR). The ASME team completed its annealing prototype demonstration in July 1996, using an indirect gas furnace at the uncompleted Public Service of Indiana's Marble Hill nuclear power plant. The MPR team's annealing prototype demonstration was scheduled to be completed in early 1997, using a direct heat electrical furnace at the uncompleted Consumers Power Company's nuclear power plant at Midland, Michigan. This paper describes the Department's annealing prototype demonstration goals and objectives; the tasks, deliverables, and results to date for each annealing prototype demonstration; and the remaining annealing technology challenges

Instrumentation and control of future sodium cooled fast reactors - Design improvements

International Nuclear Information System (INIS)

Madhusoodanan, K.; Sakthivel, M.; Chellapandi, P.

2013-06-01

India's fast reactor program started with the 40 MWt Fast Breeder Test Reactor. 500 MWe Prototype Fast Breeder Reactor (PFBR) is currently under construction at Kalpakkam. Safety of PFBR is enhanced by improved design features of I and C system. Since the design of Instrumentation and control (I and C) of PFBR, considerable improvements in terms of advancement in technology and indigenization has taken place. Further improvements in I and C is proposed for solving many of the difficulties faced during the design and construction phases of PFBR. Design improvements proposed are covered in this paper which will make the implementation and maintenance of I and C of future SFRs easier. (authors)

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Long Term Storage with Surveillance of Canadian Prototype Nuclear Power Reactors

International Nuclear Information System (INIS)

Janzen, Rick

2008-01-01

Atomic Energy of Canada (AECL) was originally formed by the government of Canada in 1952 to perform research in radiation and nuclear areas. In the mid 1950's Canada decided to limit itself to peaceful uses of nuclear energy and AECL embarked on several research and development programs, one of them being the development of nuclear power plants. This led to the development of the CANDU TM design of heavy water power reactors, of which there are now 29 operating around the world. This presentation discusses the present state of the first two CANDU TM prototype reactors and a prototype boiling light water reactor and lessons learnt after being in a long-term storage with surveillance state for more than 20 years. AECL facilities undergo decommissioning by either a prompt or a deferred removal approach. Both approaches are initiated after an operating facility has been declared redundant and gone through final operational shutdown. For the deferred approach, initial decommissioning activities are performed to put the facility into a sustainable, safe, shutdown state to minimize the hazards and costs of the ensuing extended storage with surveillance (SWS) or Safestor phase. At the appropriate time, the facility is dismantled and removed, or put into a suitable condition for re-use. AECL has a number of facilities that were built during its history, and some of these are now redundant or will become redundant in the near future. The deferred removal approach is part of AECL's decommissioning strategy for several reasons: 1. Reduction in radiation doses to workers during the final dismantling, 2. No facilities are available yet in Canada for the management of quantity of wastes arising from decommissioning, 3. Financial constraints presented by the number of facilities that will undergo decommissioning, compared to the availability of funds to carry out the work. This has led to the development of a comprehensive decommissioning plan that includes all of AECL's redundant

Passive and engineered safety features of the prototype fast reactor (PFR), Dounreay

International Nuclear Information System (INIS)

Gregory, C.V.

1991-01-01

Prototype fast reactor (PFR) combines passive and engineered safety features. Natural convection, a strong negative power coefficient, the decay heat removal system, and a fuel design able to operate beyond failure are all inherent and passive safety features of the PFR. The reliable shutdown system and the protection provided against SGU leaks are example of engineered protection. Experience at PFR demonstrates the worth and potential of a range of passive and engineered safeguards

Annual technical report of the prototype fast breeder reactor Monju. 2012

International Nuclear Information System (INIS)

2013-09-01

The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2012. From the aspect of the design evaluation, the following items are summarized: 1) Comprehensive safety assessments of Monju taking into account the accident at Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company, 2) Evaluation of nuclear characteristics based on the data of core confirmation test, 3) Evaluation of hydrogen flux from steam generator tubes, 4) Construction of the advanced safeguards system, 5) Development of a plant dynamics analytical model for the Monju ex-vessel fuel storage system. Then, from the aspect of the maintenance technology, the following items are summarized: 1) Response to the administrative order to the defect of maintenance management, 2) Recovery of in vessel transfer machine dropping accident, 3) Work management by introduction of packaged isolation, 4) Evaluation of result of vibration control of RID sampling blower for secondary sodium loop. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

The Atomics International (AI) prototype large breeder reactor (PLBR)

International Nuclear Information System (INIS)

McDonald, J.S.; Campise, A.V.; Brunings, J.

1978-01-01

The AI-PLBR breeder plant design prepared for ERDA and EPRI is of 1000 MWe size, utilizing a loop-type sodium system configuration and producing 2200 psig/850 0 F steam. A 'bullseye' core geometry type sodium system configuration and is employed with Pu0 2 -UO 2 fuel and UO 2 fertile material. The reactor outlet coolant temperature is 930 0 F. A modified 'A'-frame refueling system is employed, which is capable of handling 1/3 of a core loading in 12 days. An inducer-type mechanical pump is used in the primary circuit because of its excellent NPSH characteristics. Hockey sticks steam generators are used to produce near-fossil steam conditions at the turbine throttle. The balance of plant (BOP) design was developed by the architectural-engineering firm of Burns and Roe. It includes Allis-Chalmers - KWU 1800-rpm tandem-compound turbine, selected with high-,intermediate-, and two double-flow low-pressure cylinders equipped with two moisture separator cyclones. A design feature to enhance the licensability of the reference design is the three-level Decay Heat Removal System (DHRS), which consists of normal, backup, and diverse decay heat removal paths. The Atomics International PLBR design illustrated the technical soundness of the LMFBR system for meeting the world's long-term electrical energy supply needs. The design includes design features to assure licensability and innovative engineering features to enhance reliability, constructability, economics, and U.S. utility grid compatibility. The design concept provides a sound basis for the future detailed design of a prototype plant and subsequent development of larger LMFBR plants. (author)

Towards EPR (European pressurized reactor)

International Nuclear Information System (INIS)

Anon.

2003-01-01

According to the French industry minister, it is nonsense continuing delaying the construction of an EPR prototype because France needs it in order to renew timely its park of nuclear reactors. The renewing is expected to begin in 2020 and will be assured with third generation reactors like EPR. A quick launching of the EPR prototype is necessary to have it being in service by 2012, the feedback operating experience that will be accumulated over the 8 years that will follow will be necessary to optimize the industrial version and to have it ready by 2020. The EPR reactor has indisputable assets: modern, safer, more competitive and it will produce less wastes than present nuclear reactors. The construction cost of an EPR prototype is estimated to 3 milliard Euros and the nuclear industry operators propose to finance it completely. The EPR prototype does not jeopardize the ambitious French program about renewable energy sources, France is committed to produce 21% of its electricity from renewable energies by 2010 and 10 milliard Euros will be invested over this period on wind energy. Nuclear energy and alternative energies must be considered as 2 aspects of a diversified energy policy. (A.C.)

arXiv Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

CERN Document Server

Abreu, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B.C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L.N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

2018-05-03

The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration so...

International review on safety requirements for the prototype fast breeder reactor “Monju”

International Nuclear Information System (INIS)

2016-01-01

In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

Irradiation creep of various ferritic alloys irradiated at {approximately}400{degrees}C in the PFR and FFTF reactors

Energy Technology Data Exchange (ETDEWEB)

Toloczko, M.B.; Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Company, Richland, WA (United States)

1997-04-01

Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400{degrees}C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400{degrees}C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 x 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

ISACS-1, a limited prototype of an advanced control room

International Nuclear Information System (INIS)

Follesoe, K.; Foerdestroemmen, N.; Haugset, K.; Holmstroem, C.

1994-01-01

The concept of an Integrated Surveillance And Control System (ISACS) has been developed into a prototype, ISACS-1, which presently is in operation at the simulator-based experimental control room HAMMLAB of the OECD Halden Reactor Project. Characteristics of ISACS is that it covers the whole interface between the process and the operator, and this interface is fully computerized using tools like Cathode Ray Tubes (CRTs) and dynamic keyboards. In addition, a large number of computerized operator support systems (COSSs) are included in ISACS, assisting the operator in functions like disturbance detection and diagnosis, identification of relevant actions, and implementation of procedures. An information coordinator called ''Intelligent Coordinator'' (IC) in ISACS observes the information received from the process and the COSSs, generates new high-level information and structures and prioritizes information to be presented to the operator. The limited ISACS-1 prototype was completed in early 1991. An extensive evaluation programme is in progress. This paper will describe main features of the system and some of the conclusions to be drawn from the evaluation programme. (author). 5 refs, 2 figs

Prototype for dS/CFT correspondence

International Nuclear Information System (INIS)

Gueijosa, Alberto; Lowe, David A.; Murugan, Jeff

2005-01-01

We consider dS 2 /CFT 1 where the asymptotic symmetry group of the de Sitter spacetime contains the Virasoro algebra. We construct representations of the Virasoro algebra realized in the Fock space of a massive scalar field in de Sitter space, built as excitations of the Euclidean vacuum state. These representations are unitary, without highest weight, and have vanishing central charge. They provide a prototype for a new class of conformal field theories dual to de Sitter backgrounds in string theory. The mapping of operators in the CFT to bulk quantities is described in detail. We comment on the extension to dS 3 /CFT 2

Prototype fuel fabrication for nuclear reactors of Laguna Verde

International Nuclear Information System (INIS)

Nocetti, C.; Torres, J.; Medrano, A.

1996-01-01

Four prototype fuel bundles for the Laguna Verde Nuclear Power Plant have been fabricated. the type of nuclear fuel produced is described and the process used is commented. As an example of the fabrication criteria adopted, the production model to determine the density of the U O 2 pellets for the different batches of ceramic powder is described. the results are evaluated using the statistical indexes C p and C pk . (author)

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; El-Genk, Mohamed S.

2006-01-01

The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

State of development of high temperature gas-cooled reactors in foreign countries

International Nuclear Information System (INIS)

Sudo, Yukio

1990-01-01

Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

Human Factors and Technical Considerations for a Computerized Operator Support System Prototype

Energy Technology Data Exchange (ETDEWEB)

Ulrich, Thomas Anthony [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lew, Roger Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Medema, Heather Dawne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Boring, Ronald Laurids [Idaho National Lab. (INL), Idaho Falls, ID (United States); Thomas, Kenneth David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

A prototype computerized operator support system (COSS) has been developed in order to demonstrate the concept and provide a test bed for further research. The prototype is based on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment. A COSS demonstration scenario has been developed for the prototype involving the Chemical & Volume Control System (CVCS) of the PWR simulator. It involves a primary coolant leak outside of containment that would require tripping the reactor if not mitigated in a very short timeframe. The COSS prototype presents a series of operator screens that provide the needed information and soft controls to successfully mitigate the event.

Development of long-life neutron detectors for the prototype heavy water reactor 'Fugen'

International Nuclear Information System (INIS)

Ohteru, Shigeru; Shirayama, Shimpey.

1981-01-01

The development of long-life neutron detectors as the flux monitors for the prototype heavy water reactor has been made. Three kinds of neutron monitors, namely start-up monitor (SUM), power up monitor (PUM) and local power monitor (LPM), are provided. The LPM consists of 4 ion chamber type neutron detectors and a guide tube of power calibration monitor (PCM). This is useful for reactor control and fuel soundness monitor. The improvement of the neutron detectors was made for the operation under high neutron flux and gamma-ray heating. For the long-life operation, U-234 was mixed into U-235 for the conversion in the detectors. The ratio of U-234 to U-235 is 3 to 1. The PCM is also an ion chamber type detector with U-235. The mixing ratio of U-234 to U-235 was determined by a test with the JMTR. The characteristic performance was also investigated by the JMTR. After the completion of Fugen, various tests on the long-life detectors were performed with Fugen. It was hard to test the output linearity of the detectors with a large scale reactor. Therefore, it was tested that the operation range of the detectors is within the linear region of detector output. The voltage-current characteristics and the correlation of output current and saturation current were measured. The variation of the neutron sensitivity of the detectors with the cumulative dose was also studied. (Kato, T.)

Conceptual design of high resolution and reliable density measurement system on helical reactor FFHR-d1 and demonstration on LHD

International Nuclear Information System (INIS)

Akiyama, T.; Yasuhara, R.; Isobe, M.; Sakamoto, R.; Goto, T.; Kawahata, K.; Sagara, A.; Nakayama, K.; Okajima, S.

2014-10-01

This paper describes a conceptual design of the density measurement system on the helical reactor FFHR-d1 based on its quantitative operation scenario. The density measurement is required to meet the reactor design, and to have a high density resolution of the order of 10 17 m -3 with a time resolution of 10 ms and high reliability (no fringe jump). “A dispersion interferometer” is designed and a prototype is tested and installed on LHD, which can realize a demo relevant density plasma. The prototype demonstrates the feasibility on a demo reactor. (author)

Terrestrial Energy bets on molten salt reactors

International Nuclear Information System (INIS)

Anon.

2015-01-01

Terrestrial Energy is a Canadian enterprise, founded in 2013, for marketing the integral molten salt reactor (IMSR). A first prototype (called MSRE and with an energy output of 8 MW) was designed and operated between 1965 and 1969 by the Oak Ridge National Laboratory. IMSR is a small, modular reactor with a thermal energy output of 400 MW. According to Terrestrial Energy the technology of conventional power reactors is too complicated and too expensive. On the contrary IMSR's technology appears to be simple, easy to operate and affordable. With a staff of 30 people Terrestrial Energy appears to be a start-up in the nuclear sector. A process of pre-licensing will be launched in 2016 with the Canadian nuclear safety authority. (A.C.)

Instrumentation and control for reactor power setback in PFBR

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Vasal, Tanmay; Nagaraj, C.P.; Madhusoodanan, K.

2013-01-01

In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Kapoor, K.; Prahlad, B.

2012-01-01

Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

Textural and mechanical characterization of C-S-H gels from hydration of synthetic T1-C3S, β-C2S and their blends

Directory of Open Access Journals (Sweden)

Dolado, J. S.

2011-06-01

Full Text Available The textural and mechanical characterization of C-S-H gels formed from the hydration of pure T1-C3S, β-C2S and their blends are studied by Nitrogen sorption and nanoindentation experiments. The surface area and nanoporosity of C-S-H gels formed from the hydration of β-C2S and the 30-70 (T1-C3S and β-C2S mixture are higher than those from hydration of T1-C3S, and 70-30, with the difference decreasing with hydration age. Such changes are well supported by findings of nanoindentation study, which shows the greater relative volume of C-S-H phases with lower densities in the β-C2S and the 30-70 pastes. With the increase in hydration age, the relative volume of C-S-H phases with higher densities increased at the expenses of those with lower density. Important quantitative correlations were found among these textural characteristics and the mean chain length, determined from 29Si magic-angle-spinning (MAS NMR, of the C-S-H gels.La caracterización textural y mecánica de geles C-S-H formados a partir de la hidratación de muestras puras de T1-C3S, ß-C2S y sus mezclas ha sido estudiada por medio de adsorción de nitrógeno y nanoindentación. El área superficial y la nano-porosidad de los geles formados durante la hidratación del ß-C2S y la mezcla 30-70 (T1-C3S- ß-C2S son mayores que los correspondientes a los geles del T1-C3S, y la mezcla 70-30; esta diferencia disminuye con el tiempo de hidratación. Estos cambios coinciden con los resultados de nanoindentación que indican un aumento de volumen relativo de las fases C-S-H con una densidad menor en el caso del ß-C2S y la mezcla 30-70. Al aumentar el tiempo de hidratación, el volumen relativo de fases C-S-H de mayor densidad aumenta a expensas de aquellas de menor densidad. Importantes correlaciones cuantitativas se establecen entre las características texturales y la longitud de cadena media del gel C-S-H, determinada mediante RMN-MAS de 29Si.

EBR-2 [Experimental Breeder Reactor-2] test programs

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

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, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Programme and current status of fast breeder reactor development in Japan

International Nuclear Information System (INIS)

Suita, T.; Oyama, A.

1977-01-01

In 1967 the Japan Atomic Energy Commission revised her long term programme after a two year study for giving principles to her nuclear energy development programme, which indicated the dominant role of nuclear energy mid 1980's in the electric power generation and stressed the necessity of developing fast breeder reactors. It also recommended to organize a nucleus to undertake this nation-wide project, bringing together the total capability available throughout the country. Accordingly, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established in 1967 to develop two sodium-cooled fast reactors, an experimental fast reactor of about 100 MW thermal and a prototype fast breeder reactor of about 300 MW electrical, both using mixed oxide fuels. Construction of the experimental fast reactor started in 1970 and was essentially completed at the end of in 1974. The precommissioning test was followed in parallel with re-evaluating quality assurance of all systems. Physics test will be initiated around the end of 1976. The conceptual design of the prototype fast breeder reactor is now toward its final stage. Surveys on its proposed site have just started. Construction will start in 1978. Beside R and D works conducted by many organizations in Japan as well as under the international cooperation, several key test facilities were installed by PNC itself to conduct in-sodium test of full-size prototype components including 50 MW steam generators and post-irradiation-examination of fuels and materials. Recently an interim report was issued to an ad-hoc committee organized by JAEC to evaluate future prospect of the fuel cycle and power reactors. This recommended start of construction of the prototype reactor as scheduled and the large demonstration reactor to be followed to the prototype. Thus the fast breeder reactor is indicated as the most indispensable in 1990's

Annual technical report of the prototype fast breeder reactor Monju. 2011

International Nuclear Information System (INIS)

2012-08-01

The prototype fast breeder reactor Monju has accumulated technical achievements in order to establish the fast breeder reactor cycle technology in Japan using the operation and maintenance experience, etc. This annual report summarizes the primary achievements and the data related to the plant management in Monju during fiscal 2011. From the aspect of the design evaluation, the following items are summarized: 1) the evaluation of the decay heat removal of Monju core by natural convection, and the safety measures against earthquake and tsunami, which were carried out from the lessons learned at the Fukushima-daiichi accident due to the Great East Japan Earthquake on March 11, 2011, 2) the control rod worth confirmation and the evaluation of nuclear data library based on the data of Core Confirmation Test, which is the first step of Monju system startup test restarted in 2010, 3) the evaluation of the hydrogen concentration behavior, which detects the leak of water from the heat transfer tube of steam generator. Then, from the aspect of the maintenance technology, the following items are summarized: 1) the results of the function confirmation test on the water/steam system, after the long-term suspension, 2) confirmation of the integrity of cracked cylinder liners of emergency diesel generator, 3) replacement of the annulus ventilation duct, 4) evaluation of reduction of the periodic inspection schedule after full power operation. Furthermore, from the aspect of the plant management, this report summarizes the data related to the main topics, the history of plant condition, the sodium and water purity management, the radioactive waste management, the equipment inspection and so on. (author)

Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-06-15

The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

Textural and mechanical characterization of C-S-H gels from hydration of synthetic T1-C3S, β-C2S and their blends

International Nuclear Information System (INIS)

Goni, S.; Guerrero, A.; Puertas, F.; Hernandez, M. S.; Palacios, M.; Dolado, J. S.; Zhu, W.; Howind, T.

2011-01-01

The textural and mechanical characterization of C-S-H gels formed from the hydration of pure T1-C 3 S, β-C 2 S and their blends are studied by Nitrogen sorption and nano indentation experiments. The surface area and nano porosity of C-S-H gels formed from the hydration of β-C 2 S and the 30-70 (T1-C 3 S and β-C 2 S mixture) are higher than those from hydration of T1-C 3 S, and 70-30, with the difference decreasing with hydration age. Such changes are well supported by findings of nano indentation study, which shows the greater relative volume of C-S-H phases with lower densities in the β-C 2 S and the 30-70 pastes. With the increase in hydration age, the relative volume of C-S-H phases with higher densities increased at the expenses of those with lower density. Important quantitative correlations were found among these textural characteristics and the mean chain length, determined from 2 9Si magic-angle-spinning (MAS) NMR, of the C-S-H gels. (Author) 36 refs.

Design Feature and Prototype Testing Methodology of DHIC's Nuclear I and C System

International Nuclear Information System (INIS)

Kim, K.H.; Baeg, S.Y.; Kim, S.A.; Lee, S.J.; Yoon, S.P.; Park, C.Y.

2011-01-01

The DHIC has developed an I and C system for a nuclear power plant through a Korean Government R and D project since 2001. This I and C system was designed and implemented to be applied for the new 1400MW nuclear power plant of KHNP. This system's design is based on the class-1E PLC platform and the non-class1E DCS platform. The PPS, the ESF-CCS, the RCOPS, the QIAS-P/N, the PCS, the NPCS, the P-CCS and the NIMS were designed, implemented and tested. The R and D project has been developed under a systematic and guided QA plan, but it is not easy to be applied for a new NPP such as Shin-Ulchin 1 and 2. To resolve problems of the first-application concerns, a new idea of integrated performance testing was adopted. A main control room for a verification test facility was constructed and it has features of a compact, video-based man-machine interface. The MCR includes five operation consoles, a Large Display Panel. A test system for a verification test facility is implemented as similar as a control and protection system of SUN 1 and 2. Integration level tests such as a system test, an interface test, a MMI test, a system function/performance test, a failure mode test, a response time test, a network load test, an alarm test, a reactor power cutback system test, an unit load transient test and a scenario test were performed using the prototype test facilities. These kinds of testing can verify and pre-validate the integrated I and C system's performance and flexibility. It could offer an implementation training before construction and also minimize trial errors to be found in the site. (author)

Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

International Nuclear Information System (INIS)

Veshnyakov, K.B.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

2000-01-01

Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

Synthesis of the IRSN report on its analysis of the safety guidance package (DOrS) of the ASTRID reactor project. Safety guidance document for the ASTRID prototype: Referral to the GPR. Opinion related to the safety guidance document of the ASTRID reactor project. ASTRID prototype: Safety guidance document for the ASTRID prototype

International Nuclear Information System (INIS)

Lachaume, Jean-Luc; Niel, Jean-Christophe

2013-01-01

A first document indicates the improvement guidelines for the ASTRID project based on the French experience in the field of sodium-cooled fast neutron reactors, addresses the safety objectives as they are presented for the ASTRID project, discusses how the project includes a regulation and design referential, and how it addresses various aspects of the design approach (ranking and analysis of operation situations, defence in depth, use of probabilistic studies, safety classification and qualification to accidental situations, taking internal and external aggressions into account and taking severe accidents into account at the design level). It comments the guidelines related to the first two barriers, to main safety functions (control of reactivity and of reactor cooling, containment of radioactive and toxic materials), to dismantling, to R and D for safety support. A second document is a letter sent by the ASN to the GPR (permanent group of experts in charge of nuclear reactors) about the safety guidance document for the ASTRID prototype. The third document is the answer and contains comments and recommendations by this group about the content of this document, and therefore addresses the same topics as the first document. The last document defines the framework of the approach to this document

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

International Nuclear Information System (INIS)

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

Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

International Nuclear Information System (INIS)

Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

1991-01-01

The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

Use of self powered neutron detectors in the IEA-R1 reactor

International Nuclear Information System (INIS)

Galo Rocha, F. del.

1989-01-01

A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Chan Bock Lee

2016-10-01

Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Textural and mechanical characterization of C-S-H gels from hydration of synthetic T1-C{sub 3}S, {beta}-C{sub 2}S and their blends; Caracterizacion textural y mecanica de geles C-S-H formados en la hidratacion de muestras sinteticas T1-C{sub 3}S, {beta}-C{sub 2}S y sus mezclas

Energy Technology Data Exchange (ETDEWEB)

Goni, S.; Guerrero, A.; Puertas, F.; Hernandez, M. S.; Palacios, M.; Dolado, J. S.; Zhu, W.; Howind, T.

2011-07-01

The textural and mechanical characterization of C-S-H gels formed from the hydration of pure T1-C{sub 3}S, {beta}-C{sub 2}S and their blends are studied by Nitrogen sorption and nano indentation experiments. The surface area and nano porosity of C-S-H gels formed from the hydration of {beta}-C{sub 2}S and the 30-70 (T1-C{sub 3}S and {beta}-C{sub 2}S mixture) are higher than those from hydration of T1-C{sub 3}S, and 70-30, with the difference decreasing with hydration age. Such changes are well supported by findings of nano indentation study, which shows the greater relative volume of C-S-H phases with lower densities in the {beta}-C{sub 2}S and the 30-70 pastes. With the increase in hydration age, the relative volume of C-S-H phases with higher densities increased at the expenses of those with lower density. Important quantitative correlations were found among these textural characteristics and the mean chain length, determined from {sup 2}9Si magic-angle-spinning (MAS) NMR, of the C-S-H gels. (Author) 36 refs.

Instrumentation and control improvements at Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Christensen, L.J.; Planchon, H.P.

1993-01-01

The purpose of this paper is to describe instrumentation and control (I ampersand C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I ampersand C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I ampersand C systems of the next generation of liquid metal reactor (LMR) plants

Fabrication and quality assurance of some important components and sub-assemblies for Prototype Fast Breeder Reactor (PFBR) project

International Nuclear Information System (INIS)

Dutta, N.G.; More, S.S.

2010-01-01

Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500 MWe prototype fast breeder reactor (PFBR) at Kalpakkam, Chennai. In this very important and prestigious national programmed M/s Kay Bouvet Engg. Pvt. Ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies. M/s KBEPL is engaged in manufacturing, quality assurance and supply of many subassemblies of PFBR like under water trolley (UWT), shielding door, container and container storage rack (CSR), vessel in fuel transfer cell (FTC), personnel air lock (PAL), emergency air lock (EAL) and material air lock (MAL), absorber rod drive mechanism (ARDM) flask assembly and carriage in MAL etc. Two partition doors and four nos. of embedded parts (SS 304L) have already been supplied to Bhavini. The paper deals with manufacturing and Q.A. activities being carried out for supply of these important assemblies to PFBR projects. (author)

Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

International Nuclear Information System (INIS)

Mouri, Tomoaki; Kunitomi, Kazuhiko

2008-01-01

A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

2017-06-15

Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

International Nuclear Information System (INIS)

Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

2013-01-01

Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

International review on safety requirements for the prototype fast breeder reactor “Monju” (Translated document)

International Nuclear Information System (INIS)

2016-02-01

In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

Genetic analysis of complement C1s deficiency associated with systemic lupus erythematosus highlights alternative splicing of normal C1s gene

DEFF Research Database (Denmark)

Amano, Mariane T; Ferriani, Virgínia P L; Florido, Marlene P C

2008-01-01

Deficiencies of complement proteins of the classical pathway are strongly associated with the development of autoimmune diseases. Deficiency of C1r has been observed to occur concomitantly with deficiency in C1s and 9 out of 15 reported cases presented systemic lupus erythematosus (SLE). Here, we...... describe a family in which all four children are deficient in C1s but only two of them developed SLE. Hemolytic activity mediated by the alternative and the lectin pathways were normal, but classical pathway activation was absent in all children's sera. C1s was undetectable, while in the parents' sera...

U.S. Sodium Fast Reactor Codes and Methods: Current Capabilities and Path Forward

Energy Technology Data Exchange (ETDEWEB)

Brunett, A. J.; Fanning, T. H.

2017-06-26

The United States has extensive experience with the design, construction, and operation of sodium cooled fast reactors (SFRs) over the last six decades. Despite the closure of various facilities, the U.S. continues to dedicate research and development (R&D) efforts to the design of innovative experimental, prototype, and commercial facilities. Accordingly, in support of the rich operating history and ongoing design efforts, the U.S. has been developing and maintaining a series of tools with capabilities that envelope all facets of SFR design and safety analyses. This paper provides an overview of the current U.S. SFR analysis toolset, including codes such as SAS4A/SASSYS-1, MC2-3, SE2-ANL, PERSENT, NUBOW-3D, and LIFE-METAL, as well as the higher-fidelity tools (e.g. PROTEUS) being integrated into the toolset. Current capabilities of the codes are described and key ongoing development efforts are highlighted for some codes.

Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

International Nuclear Information System (INIS)

Lerner, Ana Maria; Madariaga, Marcelo

2002-01-01

The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

Moving ring reactor 'Karin-1'

International Nuclear Information System (INIS)

1983-12-01

The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

International Nuclear Information System (INIS)

1984-06-01

The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

Performance tests of a full scale prototype of the Belle II TOP counter with cosmic muons and 2.1 GeV/c positron beam

Energy Technology Data Exchange (ETDEWEB)

Hayakawa, Tomokatsu

2014-12-01

A time-of-propagation detector named TOP is a hadronic particle identification system for the Belle II experiment. We have produced a full scale prototype of the Belle II TOP counter and tested with cosmic muons at KEK and the 2.1 GeV/c positrons at SPring-8 LEPS beamline. The procedures for the quartz acceptance test and assembly worked well and the first quartz radiator was successfully fabricated. The obtained test data shows good agreement with Monte Carlo expectation. - Highlights: • We have successfully produced a full scale prototype of the Belle II TOP counter. • The prototype counter has been tested at KEK and SPring-8 LEPS beamline. • The obtained test data shows good agreement with Monte Carlo expectation.

Human genes for complement components C1r and C1s in a close tail-to-tail arrangement

International Nuclear Information System (INIS)

Kusumoto, H.; Hirosawa, S.; Salier, J.P.; Hagen, F.S.; Kurachi, K.

1988-01-01

Complementary DNA clones for human C1s were isolated from cDNA libraries that were prepared with poly(A) + RNAs of human liver and HepG2 cells. A clone with the largest cDNA insert of 2,664 base pairs (bp) was analyzed for its complete nucleotide sequence. It contained 202 bp of a 5' untranslated region, 45 bp of coding for a signal peptide (15 amino acid residues), 2,019 bp for complement component C1s zymogen (673 amino acid residues), 378 bp for a 3' untranslated region, a stop codon, and 17 bp of a poly(A) tail. The amino acid sequence of C1s was 40.5% identical to that of C1r, with excellent matches of tentative disulfide bond locations conserving the overall domain structure of C1r. DNA blotting and sequencing analyses of genomic DNA and of an isolated genomic DNA clone clearly showed that the human genes for C1r and C1s are closely located in a tail-to-tail arrangement at a distance of about 9.5 kilobases. Furthermore, RNA blot analyses showed that both C1r and C1s genes are primarily expressed in liver, whereas most other tissues expressed both C1r and C1s genes at much lower levels (less than 10% of that in liver). Multiple molecular sizes of specific mRNAs were observed in the RNA blot analyses for both C1r and C1s, indicating that alternative RNA processing(s), likely an alternative polyadenylylation, might take place for both genes

Performance evaluation of domestic prototype dose area product meter SFT-1

International Nuclear Information System (INIS)

Lee, Ho Sun; Han, Seong Gyu; Roh, Young Roh; Lim, Hyun Jong; Kim, Jung Min; Kim, Jong Uk; Chae, Hyun Sik; Yoon, Yong Su

2016-01-01

The importance of radiation dose display of medical X-ray equipment was emphasized, while third edition of IEC(International Electrotechnical Commission) 60601 started to apply. The existing medical X-ray equipment selected a method for attaching the DAP(Dose Area Product) meter when the dose display. However, because the DAP meter was dependent on all of the income, And it did not yet produced in Korea. So, we received the support of Seoul R and BD Program(Grants No. C1152055) to produce DAP meter prototype of the Domestically technology. In this study, the performance of this prototype was evaluated by comparing the German company's product. Evaluation item was an electronic capture performance, radiation dose dependence, radiation quality dependence, energy transmittance, repeatability, light transmittance of 6 entries. And IEC 60580 was based on this evaluation. Evaluation results were electronic capture performance intrinsic error 9.5%, radiation dose dependence limits of variation 1%, repeatabilit y coefficient of variation 2%, energy transmittance 91% each assessment was passed. However radiation quality dependence limits of variation 29%, light transmittance 55% was less than acceptance criteria

Measurement of the $\\eta_c (1S)$ production cross-section in proton-proton collisions via the decay $\\eta_c (1S) \\rightarrow p \\bar{p}$

CERN Document Server

Aaij, Roel; Adeva, Bernardo; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreassen, Rolf; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Belogurov, Sergey; Belous, Konstantin; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Bird, Thomas; Bizzeti, Andrea; Bjørnstad, Pål Marius; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borgia, Alessandra; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Brambach, Tobias; Bressieux, Joël; Brett, David; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Brown, Henry; Bursche, Albert; Busetto, Giovanni; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Ciba, Krzystof; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cogoni, Violetta; Cojocariu, Lucian; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Counts, Ian; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dalseno, Jeremy; David, Pascal; David, Pieter; Davis, Adam; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Silva, Weeraddana; De Simone, Patrizia; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Di Canto, Angelo; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dossett, David; Dovbnya, Anatoliy; Dreimanis, Karlis; Dujany, Giulio; Dupertuis, Frederic; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egede, Ulrik; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Graverini, Elena; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farinelli, Chiara; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fol, Philip; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Francisco, Oscar; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garofoli, Justin; Garra Tico, Jordi; Garrido, Lluis; Gaspar, Clara; Gauld, Rhorry; Gavardi, Laura; Gavrilov, Gennadii; Geraci, Angelo; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianelle, Alessio; Gianì, Sebastiana; Gibson, Valerie; Giubega, Lavinia-Helena; Gligorov, Vladimir; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Gregson, Sam; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Hampson, Thomas; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Hunt, Philip; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jaton, Pierre; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kaballo, Michael; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kelsey, Matthew; Kenyon, Ian; Ketel, Tjeerd; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Korolev, Mikhail; Kozlinskiy, Alexandr; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kurek, Krzysztof; Kvaratskheliya, Tengiz; La Thi, Viet Nga; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lambert, Robert W; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Leo, Sabato; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Lohn, Stefan; Longstaff, Iain; Lopes, Jose; Lopez-March, Neus; Lowdon, Peter; Lucchesi, Donatella; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Machefert, Frederic; Machikhiliyan, Irina V; Maciuc, Florin; Maev, Oleg; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Märki, Raphael; Marks, Jörg; Martellotti, Giuseppe; Martens, Aurelien; Martín Sánchez, Alexandra; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathe, Zoltan; Matteuzzi, Clara; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; McSkelly, Ben; Meadows, Brian; Meier, Frank; Meissner, Marco; Merk, Marcel; Milanes, Diego Alejandro; Minard, Marie-Noelle; Moggi, Niccolò; Molina Rodriguez, Josue; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Katharina; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Nicol, Michelle; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Oggero, Serena; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Orlandea, Marius; Otalora Goicochea, Juan Martin; Owen, Patrick; Oyanguren, Maria Arantza; Pal, Bilas Kanti; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parkes, Christopher; Parkinson, Christopher John; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pazos Alvarez, Antonio; Pearce, Alex; Pellegrino, Antonio; Pepe Altarelli, Monica; Perazzini, Stefano; Perez Trigo, Eliseo; Perret, Pascal; Perrin-Terrin, Mathieu; Pescatore, Luca; Pesen, Erhan; Petridis, Konstantin; Petrolini, Alessandro; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Playfer, Stephen; Plo Casasus, Maximo; Polci, Francesco; Poluektov, Anton; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Rachwal, Bartolomiej; Rademacker, Jonas; Rakotomiaramanana, Barinjaka; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruiz, Hugo; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrie, Mauro; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Seco, Marcos; Semennikov, Alexander; Sepp, Indrek; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Silva Coutinho, Rafael; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Anthony; Smith, Edmund; Smith, Eluned; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Sparkes, Ailsa; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Stroili, Roberto; Subbiah, Vijay Kartik; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szczypka, Paul; Szilard, Daniela; Szumlak, Tomasz; T'Jampens, Stephane; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Tolk, Siim; Tomassetti, Luca; Tonelli, Diego; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Tran, Minh Tâm; Tresch, Marco; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ubeda Garcia, Mario; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; Voss, Helge; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Whitehead, Mark; Wicht, Jean; Wiedner, Dirk; Wilkinson, Guy; Williams, Matthew; Williams, Mike; Wilschut, Hans; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xing, Zhou; Xu, Zhirui; Yang, Zhenwei; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Wen Chao; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang; Zvyagin, Alexander

2015-01-01

The production of the $\\eta_c (1S)$ state in proton-proton collisions is probed via its decay to the $p \\bar{p}$ final state with the LHCb detector, in the rapidity range $2.0 6.5$ GeV/c. The cross-section for prompt production of $\\eta_c (1S)$ mesons relative to the prompt $J/\\psi$ cross-section is measured, for the first time, to be $\\sigma_{\\eta_c (1S)}/\\sigma_{J/\\psi} = 1.74 \\pm 0.29 \\pm 0.28 \\pm 0.18 _{B}$ at a centre-of-mass energy $\\sqrt{s} = 7$ TeV using data corresponding to an integrated luminosity of 0.7 fb$^{-1}$, and $\\sigma_{\\eta_c (1S)}/\\sigma_{J/\\psi} = 1.60 \\pm 0.29 \\pm 0.25 \\pm 0.17 _{B}$ at $\\sqrt{s} = 8$ TeV using 2.0 fb$^{-1}$. The uncertainties quoted are, in order, statistical, systematic, and that on the ratio of branching fractions of the $\\eta_c (1S)$ and $J/\\psi$ decays to the $p \\bar{p}$ final state. In addition, the inclusive branching fraction of $b$-hadron decays into $\\eta_c (1S)$ mesons is measured, for the first time, to be $B ( b \\rightarrow \\eta_c X ) = (4.88 \\pm 0.64 \\pm ...

Gas-cooled reactors

International Nuclear Information System (INIS)

Schulten, R.; Trauger, D.B.

1976-01-01

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

International Nuclear Information System (INIS)

Journeau, Ch.

2008-01-01

The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

Core designs for new VVER reactors and operational experience of immediate prototypes

International Nuclear Information System (INIS)

Vasilchenko, I.; Mokhov, V.; Ryzhov, S.

2011-01-01

The paper covers the recent improvements analyzed in order to implement the enhanced core performances. AES-2006 reactor core design is considered from the point of view of its application and improvement in the planned VVER-TOI project and of the possibilities of using the basic engineering solutions for the cores with spectral control. The discussion of several types of mixing grids considered in the paper involves a preliminary assessment of their efficiency and the information on their introduction into pilot operation at the VVER-1000 Units. Special attention is given to the results of the operation of immediate prototypes (TVS-2 and TVS-2M) that corroborate the reliability of the design both with regard for the core geometrical stability and fuel cladding tightness

Experience of secondary cooling system modification at prototype fast breeder reactor MONJU (Translated document)

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Sakamoto, Yoshihiko

2010-09-01

The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident that occurred in December 1995. After the accident, an investigation into the cause and a comprehensive safety review of the plant were conducted, and various countermeasures for sodium leak were examined. Modification work commenced in September 2005. Since sodium, a chemically active material, is used as coolant in MONJU, the modification work required work methods suitable for the handling of sodium. From this perspective, the use of a plastic bag when opening the sodium boundary, oxygen concentration control in a plastic bag, slightly-positive pressure control of cover gas in the systems, pressing and cutting with a roller cutter to prevent the incorporation of metal fillings, etc. were adopted, with careful consideration given to experience and findings from previous modification work at the experimental fast reactor JOYO and plants abroad. Owing to these work methods, the modification work proceeded close to schedule without incident. (author)

Prototypical consolidation demonstration project - Final fuel recommendation report

International Nuclear Information System (INIS)

Piscitella, R.R.; Paskey, W.R.

1987-01-01

The Prototypical Consolidation Demonstration (PCD) Project will, in its final phase, conduct a demonstration of the equipment's ability to consolidate actual spent commercial fuel. Since budget and schedule limitations do not allow this demonstration to include all types of fuel assemblies, a selection process was utilized to identify the fuel types that would represent predominate fuel inventories and that would demonstrate the equipment's abilities. The Pressurized Water Reactor (PWR) fuel assemblies that were suggested for use in the PCD Project Hot Demonstration were Babcock and Wilcox (B and W) 15 x 15's, and Westinghouse (WE) 15 x 15's. The Boiling Water Reactor (BWR) fuel suggested was the General Electric (GE) 8 x 8

History of fast reactor development in U.S.A.-I

International Nuclear Information System (INIS)

Ninokata, Hisashi; Sasao, Nobuyki

2007-01-01

History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

International Nuclear Information System (INIS)

Tikhomirov, L.N.

1995-01-01

The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

New instrumentation for the IPR-R1 reactor of CDTN

International Nuclear Information System (INIS)

Carvalho, P.V.R. de.

1992-01-01

The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

International Nuclear Information System (INIS)

1978-01-01

Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

Enhancements to ASHRAE Standard 90.1 Prototype Building Models

Energy Technology Data Exchange (ETDEWEB)

Goel, Supriya; Athalye, Rahul A.; Wang, Weimin; Zhang, Jian; Rosenberg, Michael I.; Xie, YuLong; Hart, Philip R.; Mendon, Vrushali V.

2014-04-16

This report focuses on enhancements to prototype building models used to determine the energy impact of various versions of ANSI/ASHRAE/IES Standard 90.1. Since the last publication of the prototype building models, PNNL has made numerous enhancements to the original prototype models compliant with the 2004, 2007, and 2010 editions of Standard 90.1. Those enhancements are described here and were made for several reasons: (1) to change or improve prototype design assumptions; (2) to improve the simulation accuracy; (3) to improve the simulation infrastructure; and (4) to add additional detail to the models needed to capture certain energy impacts from Standard 90.1 improvements. These enhancements impact simulated prototype energy use, and consequently impact the savings estimated from edition to edition of Standard 90.1.

U.S. and foreign breeder reactors

International Nuclear Information System (INIS)

Hill, E.H.

1977-01-01

The running battle between Congress and the Administration over the Clinch River Breeder Reactor Plant (CRBRP) Project has provoked an increased interest in domestic and foreign breeder reactor programs. Perhaps an understanding of the history of breeders here and abroad will serve to place the CRBRP in perspective and allow some analysis of how the U.S. appears on the global canvas. Breeder reactor technology has, for the most part, settled down to concentration on the liquid metal fast breeder reactor (LMFBR). This is the result of 32 years of experience with reactors employing a fast neutron flux and even longer experience with liquid metal coolants. However, a number of U.S. utilities are sponsoring a gas cooled fast reactor program as an alternative technology to the LMFBR. This development program is supported by the U.S. Department of Energy

A Small Modular Reactor Design for Multiple Energy Applications: HTR50S

Energy Technology Data Exchange (ETDEWEB)

Yan, X.; Tachibana, Y.; Ohashi, H.; Sato, H.; Tazawa, Y.; Kunitomi, K. [Japan Atomic Energy Agency, Ibaraki (Japan)

2013-06-15

HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950 .deg. C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750 .deg. C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900 .deg. C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

Sodium heat pipe module test for the SAFE-30 reactor prototype

International Nuclear Information System (INIS)

Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

2001-01-01

Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

Public acceptance of fusion energy and scientific feasibility of a fusion reactor. DREAM (DRastically EAsy Maintenance) tokamak

International Nuclear Information System (INIS)

Nishio, Satoshi

1998-01-01

If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)

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.

Presentation Trainer Prototype 1.0

NARCIS (Netherlands)

Börner, Dirk; Specht, Marcus

2014-01-01

This software sketch comprises the first prototype of the presentation trainer. The application uses the Microsoft Kinect sensor and was built using the Processing 1.5.1 development environment. Available under the GNU LGPL licence version 3 or higher.

Real Time Computer for Plugging Indicator Control of Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Manimaran, M.; Manoj, P.; Shanmugam, A.; Murali, N.; Satya Murty, S.A.V.

2013-06-01

Prototype Fast Breeder Reactor (PFBR) is in the advanced stage of construction at Kalpakkam, India. Liquid sodium is used as coolant to transfer the heat produced in the reactor core to steam water circuit. Impurities present in the sodium are removed using purification circuit. Plugging indicator is a device used to measure the purity of the sodium. Versa Module Europa bus based Real Time Computer (RTC) system is used for plugging indicator control. Hot standby architecture consisting of dual redundant RTC system with switch over logic system is the configuration adopted to achieve fault tolerance. Plugging indicator can be controlled in two modes namely continuous and discontinuous mode. Software based Proportional-Integral-Derivative (PID) algorithms are developed for plugging indicator control wherein the set point changes dynamically for every scan interval of the RTC system. Set points and PID constants are kept as configurable in runtime in order to control the process in very efficient manner, which calls for reliable communication between RTC system and control station, hence TCP/IP protocol is adopted. Performance of the RTC system for plugging indicator control was thoroughly studied in the laboratory by simulating the inputs and monitored the control outputs. The control outputs were also monitored for different PID constants. Continuous and discontinuous mode plots were generated. (authors)

Fuel recycling and 4. generation reactors

International Nuclear Information System (INIS)

Devezeaux de Lavergne, J.G.; Gauche, F.; Mathonniere, G.

2012-01-01

The 4. generation reactors meet the demand for sustainability of nuclear power through the saving of the natural resources, the minimization of the volume of wastes, a high safety standard and a high reliability. In the framework of the GIF (Generation 4. International Forum) France has decided to study the sodium-cooled fast reactor. Fast reactors have the capacity to recycle plutonium efficiently and to burn actinides. The long history of reprocessing-recycling of spent fuels in France is an asset. A prototype reactor named ASTRID could be entered into operation in 2020. This article presents the research program on the sodium-cooled fast reactor, gives the status of the ASTRID project and present the scenario of the progressive implementation of 4. generation reactors in the French reactor fleet. (A.C.)

Light Water Reactor Sustainability Program, U.S. Efforts in Support of Examinations at Fukushima Daiichi-2017 Evaluations

Energy Technology Data Exchange (ETDEWEB)

Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-08-01

Although the accident signatures from each unit at the Fukushima Daiichi Nuclear Power Station (NPS) [Daiichi] differ, much is not known about the end-state of core materials within these units. Some of this uncertainty can be attributed to a lack of information related to cooling system operation and cooling water injection. There is also uncertainty in our understanding of phenomena affecting: a) in-vessel core damage progression during severe accidents in boiling water reactors (BWRs), and b) accident progression after vessel failure (ex-vessel progression) for BWRs and Pressurized Water Reactors (PWRs). These uncertainties arise due to limited full scale prototypic data. Similar to what occurred after the accident at Three Mile Island Unit 2, these Daiichi units offer the international community a means to reduce such uncertainties by obtaining prototypic data from multiple full-scale BWR severe accidents. Information obtained from Daiichi is required to inform Decontamination and Decommissioning activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This document, which has been updated to include FY2017 information, summarizes results from U.S. efforts to use information obtained by TEPCO Holdings to enhance the safety of existing and future nuclear power plant designs. This effort, which was initiated in 2014 by the Reactor Safety Technologies Pathway of the Department of Energy Office of Nuclear Energy Light Water Reactor (LWR) Sustainability Program, consists of a group of U.S. experts in LWR safety and plant operations that have identified examination needs and are evaluating TEPCO Holdings information from Daiichi that address these needs. Each year, annual reports include examples demonstrating that significant safety insights are being obtained in the areas of component performance, fission

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

Energy Technology Data Exchange (ETDEWEB)

Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O₂ cannot be ignored, especially for the FHR, in which environment the product, SiO₂, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

The first C4-functionalisation of condensed tannins. Phlobatannins as prototype of a new class of C-ring isomerised oligomers

International Nuclear Information System (INIS)

Steenkamp, J.A.

1986-06-01

This thesis comprises besides the characterisation of new oligomeric flavenoids from the core wood of the indigeneous Colophosphermum mopane, an investigation into the C 4 -functionalisation of flavan-3-ol analogues. The first peltogynoid biflavenoid and prototypes of a new series C-ring isomerised condensed tannins, namely the phlobatannins, were isolated. Besides 1 H- nmr-parameters for structure analysis, the natural phlobatannins were characterised and the unique ring isomerisation was investigated

Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

Directory of Open Access Journals (Sweden)

Jaewoon Yoo

2016-10-01

Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

International Nuclear Information System (INIS)

Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

2016-01-01

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

Interaction of C1q and mannan-binding lectin (MBL) with C1r, C1s, MBL-associated serine proteases 1 and 2, and the MBL-associated protein MAp19

DEFF Research Database (Denmark)

Thiel, S; Petersen, Steen Vang; Vorup-Jensen, T

2000-01-01

. There is controversy as to whether MBL can utilize C1r and C1s or, inversely, whether C1q can utilize MASP-1 and 2. Serum deficient in C1r produced no complement activation in IgG-coated microwells, whereas activation was seen in mannan-coated microwells. In serum, C1r and C1s were found to be associated only with C1q...

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Construction of a sTGC Prototype for the ATLAS-MUON Upgrade

CERN Document Server

Guan, Liang; The ATLAS collaboration; Zhu, Junjie; Smakhtin, Vladimir; Shoa, Meir

2014-01-01

The innermost station (Small Wheel) of the ATLAS muon spectrometer end-cap will be replaced with the New Small Wheel (NSW) to profit from the high luminosity runs of LHC after phase I upgrade . sTGC will be the primary trigger detectors proving Level-1 trigger as well as complementing the precision muon tracking. In order to qualify materials and gain experiences for serious mass production of NSW sTGC detector modules, a 1.3 m x 1.1 mm sTGC quadruplet prototype is successfully constructed and tested. Details of the construction procedures will be presented. Control of the module flatness, machined strip board precision and alignment of different detector layers will be shown.

U.S. progress in the development of very high density low enrichment research reactor fuels

International Nuclear Information System (INIS)

Meyer, M. K.; Wachs, D. M.; Jue, J.-F.; Keiser, D. D.; Gan, J.; Rice, F.; Robinson, A.; Woolstenhulme, N. E.; Medvedev, P.; Hofman, G. L.; Kim, Y.-S.

2012-01-01

The effort to develop low-enriched fuels for high power research reactors began world-wide in 1996. Since that time, hundreds of fuel specimens have been tested to investigate the operational limits of many variations of U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel development program has focused on the development of monolithic fuel, and is currently transitioning from conducting research experiments to the demonstration of large scale, prototypic element assemblies. These larger scale, integral fuel performance demonstrations include the AFIP-7 test of full-sized, curved plates configured as an element, the RERTR-FE irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor specific Design Demonstration Experiments, and a multi-element Base Fuel Demonstration. These tests are conducted alongside mini-plate tests designed to prove fuel stability over a wide range of operating conditions. Along with irradiation testing, work on collecting data on fuel plate mechanical integrity, thermal conductivity, fission product release, and microstructural stability is underway. (authors)

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

Assessment of weld joints of steam generator of prototype fast breeder reactor by microfocal radiography

International Nuclear Information System (INIS)

Venkatraman, B.; Saravanan, T.; Jayakumar, T.; Kalyanasundaram, P.; Raj, B.

2004-01-01

The tube to tubesheet (TTS) welds of steam generator of Prototype Fast Breeder Reactor (PFBR) are quite critical. Sodium flows on shell side and water on tube side. Any failure would thus be catastrophic. Apart from defects such as porosities, wall thinning due to concavity is endemic in such joints and needs to be detected. This paper presents the methodologies developed for quantitative evaluation of defects including wall thinning due to concavity in the TTS welds by micro focal radiography. The method has been successfully adopted in the shop floor for the evaluation of TTS welds of steam generator and evaporator. (author)

Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Bhaduri, A.K.; Indira, R.; Albert, S.K.; Rao, B.P.S.; Jain, S.C.; Asokkumar, S.

2004-01-01

Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

Energy Technology Data Exchange (ETDEWEB)

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

Gas and water permeability of concrete for reactor buildings--prototype scale specimens

International Nuclear Information System (INIS)

Mills, R.H.

1987-02-01

The permeability testing was performed on four concrete cylinders, 0.25 m in diameter and 2 m long, modelling the wall-thickness of reactor containment structures on the prototype scale. Tests were performed on the cylinders before and after artificial induction of longitudinal cracks, intented to model defects developing after some period of adverse service conditions. Permeability increased greatly with the introduction of longitudinal cracks in the concrete, and was also affected by moisture content and casting direction. The influence of reinforcing steel could not be resolved within the bounds of experimental variability. Ultrasound measurements were taken on each cylinder before and after cracking, and a correlation between increased permeability and lowered Ultrasonic Pulse Velocity was observed. Ultrasonic Pulse Velocity measurements thus show promise as a means of continuous monitoring of the integrity of the concrete barrier in service

Prometheus Reactor I&C Software Development Methodology, for Action

Energy Technology Data Exchange (ETDEWEB)

T. Hamilton

2005-07-30

The purpose of this letter is to submit the Reactor Instrumentation and Control (I&C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I&C Software Development Process Manual and Reactor Module Software Development Plan to NR for information.

Reactor shutdown system of prototype fast breeder reactor

International Nuclear Information System (INIS)

Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

2002-01-01

Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6 / ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

International Nuclear Information System (INIS)

1989-04-01

This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

Thermomechanical processing of Nb-1Zr-0.1C alloy for use in compact high temperature reactors: a first report

International Nuclear Information System (INIS)

Chakravartty, J.K.; Kapoor, R.; Suri, A.K.

2011-08-01

Nb-1Zr-0.1C is a potential material for use in high temperature nuclear reactors. Use of this alloy in components requires appropriate thermomechanical processing to break the cast microstructure and to obtain uniformly distributed fine stable precipitates so as to produce the desired mechanical properties at the high operating temperatures. This report reviews the thermomechanical processing of Nb-1Zr-0.1C alloy carried out over the years by other researchers and the high temperature creep behavior of the alloy. The hot deformation of Nb-1Zr-0.1C alloy carried out at Mechanical Metallurgy Division is also presented here. From this review it is evident that most primary hot working studies were carried out between 1500 to 1700 degC. The subsequent annealing treatments, which require holding at lower temperatures of about 1100 to 1300 degC for very long times help further transform the precipitates from coarse orthorhombic to very fine cubic. Our studies on Nb-1Zr-0.1C alloy also confirm that optimum hot working lies at temperatures beyond 1500 degC where dynamic recrystallization initiates, and optimally around 1700 degC where dynamic recrystallization transforms the microstructure. Working at temperatures lower than 1000 degC may lead to the undesirable effect of both micro as well as macro strain localization, and should be avoided. (author)

Project management plan for Reactor 105-C Interim Safe Storage project

International Nuclear Information System (INIS)

Plagge, H.A.

1996-09-01

Reactor 105-C (located on the Hanford Site in Richland, Washington) will be placed into an interim safe storage condition such that (1) interim inspection can be limited to a 5-year frequency; (2) containment ensures that releases to the environmental are not credible under design basis conditions; and (3) final safe storage configuration shall not preclude or significantly increase the cost for any decommissioning alternatives for the reactor assembly.This project management plan establishes plans, organizational responsibilities, control systems, and procedures for managing the execution of Reactor 105-C interim safe storage activities to meet programmatic requirements within authorized funding and approved schedules

U.S. Domestic Reactor Conversion Programs

International Nuclear Information System (INIS)

Woolstenhulme, Eric

2008-01-01

The Conversion Projects Include: the revision of the facilities safety basis documents and supporting analysis, the fabrication of new LEU fuel, the change-out of the reactor core, and the removal of the used HEU fuel (by INL University Fuels Program or DOE-NE). The major entities involved are: the U.S. Nuclear Regulatory Commission, the University reactor department, the fuel and hardware fabricators, the Spent fuel receipt facilities, the Spent fuel shipping services, and the U.S. Department of Energy and their subcontractors. Three major Reactor Conversion Program milestones have been accomplished since 2006: the conversion of the TRIGA reactor at Texas A and M University Nuclear Science Center, the conversion of the University of Florida Training Reactor, and the conversion of the Purdue University Reactor. Four Reactor Conversion Program milestones yet to be accomplished in 2008 and 2009: the Washington State University Nuclear Radiation Center reactor, the Oregon State University TRIGA Reactor, the University of Wisconsin Nuclear Reactor, and the Neutron Radiography Reactor Facility. NNSA is committed to doing things cheaper, better, smarter, safer through a 'Lessons Learned' process. The conversion team assessed each major activity grouping: Project Initiation, Conversion Proposal Development, Fuel Fabrication and Hardware, Core Conversion, and Spent Nuclear Fuel Removal. Issues were identified and recommendations were given

Study of the ηc(1 1S0) state of charmonium formed in bar pp annihilations and a search for the ηc'(2 1S0)

International Nuclear Information System (INIS)

Armstrong, T.A.; Bettoni, D.; Bharadwaj, V.; Biino, C.; Blanford, G.; Borreani, G.; Broemmelsiek, D.; Buzzo, A.; Calabrese, R.; Ceccucci, A.; Cester, R.; Church, M.; Dalpiaz, P.; Dalpiaz, P.F.; Dimitroyannis, D.; Fabbri, M.; Fast, J.; Gianoli, A.; Ginsburg, C.M.; Gollwitzer, K.; Govi, G.; Hahn, A.; Hasan, M.; Hsueh, S.; Lewis, R.; Luppi, E.; Macri, M.; Majewska, A.M.; Mandelkern, M.; Marchetto, F.; Marinelli, M.; Marques, J.; Marsh, W.; Martini, M.; Masuzawa, M.; Menichetti, E.; Migliori, A.; Mussa, R.; Palestini, S.; Pallavicini, M.; Passaggio, S.; Pastrone, N.; Patrignani, C.; Peoples, J. Jr.; Petrucci, F.; Pia, M.G.; Pordes, S.; Rapidis, P.; Ray, R.; Reid, J.; Rinaudo, G.; Roccuzzo, B.; Rosen, J.; Santroni, A.; Sarmiento, M.; Savrre, M.; Schultz, J.; Seth, K.K.; Smith, A.; Smith, G.A.; Sozzi, M.; Trokenheim, S.; Weber, M.F.; Werkema, S.; Zhang, Y.; Zhao, J.; Zioulas, G.

1995-01-01

The E760 Collaboration performed an experiment in the Antiproton Accumulator at Fermilab to study the two photon decay of the η c (1 1 S 0 ) charmonium state formed in bar pp annihilations. This resulted in a new measurement of the mass M ηc =2988.3 -3.1 +3.3 MeV/c 2 and of the product B(η c → bar pp)xΓ(η c →γγ) =(8.1 -2.0 +2.9 ) eV. We performed a search for the process bar pp→η c ' (2 1 S 0 )→γγ over a limited range of center-of-mass energies. Since no signal was observed, we derived upper limits on the product of branching ratios B(η c ' → bar pp)xB(η c ' →γγ) in the center-of-mass energy range 3584≤ √s ≤3624 MeV. We observed no signal for the nonresonant process bar p+p→γ+γ and obtain upper limits

HMB-45, S-100, NK1/C3, and MART-1 in metastatic melanoma.

Science.gov (United States)

Zubovits, Judit; Buzney, Elizabeth; Yu, Lawrence; Duncan, Lyn M

2004-02-01

The diagnosis of melanoma metastatic to lymph node remains a difficult problem given its histological diversity. We examined the staining patterns of S-100, NK1/C3, HMB-45, and MART-1 (DC10) in melanoma metastases to lymph nodes. Immunohistochemical stains were performed on tissue sections of 126 formalin-fixed lymph nodes from 126 patients with an established diagnosis of metastatic melanoma. A total of 98% of cases (123 of 126) stained positive for S-100, 93% (117 of 125) stained positive for NK1/C3, 82% (103 of 126) stained positive for MART-1, and 76% (95 of 125) stained positive for HMB-45. The distribution and intensity of staining varied among these markers. A diffuse staining pattern, defined as >50% of tumor cells stained, was observed in 83% of MART-1-positive cases but in only 56% of S-100-positive cases, 48% of NK1/C3-positive cases, and 34% of HMB-45-positive cases. A maximally intense signal was almost always observed for MART-1 (83% of positive cases) but was rarely observed for NK1/C3 (20%). S-100 and HMB-45 showed maximally intense staining in 50% and 54% of cases, respectively. S-100 and NK1/C3 stained both histiocytes and melanocytes, whereas MART-1 and HMB-45 stained only melanocytes. Seventy-eight cases (63%) stained positive for all 4 markers, 17 cases (14%) stained for all markers except HMB-45, 13 cases (10%) stained for all markers except MART-1, 6 cases (5%) stained only with S-100 and NK1/C3, 4 cases (3%) stained only with S-100 and HMB-45, and 2 cases stained for all markers except S-100. One case each stained for the following: only S-100, only S-100 and HMB-45, and all markers except NK1/C3. One case exhibited absence of staining for any of these markers. We demonstrate that lymph node metastases of melanoma are heterogeneous with regard to tumor marker expression. S-100 and NK1/C3 were the most sensitive stains for detecting metastatic melanoma; however, they both also stain other nontumor cells in lymph nodes. MART-1 did not stain

United Kingdom and USSR reactor types

International Nuclear Information System (INIS)

Lewins, Jeffery

1988-01-01

The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

Recent U.S. reactor operating experience

International Nuclear Information System (INIS)

Stello, V. Jr.

1977-01-01

A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

The low power miniature neutron source reactors: Design, safety and applications

International Nuclear Information System (INIS)

Ahmed, Y.A.; Ewa, I.O.B.; Umar, M.; Bezboruah, T.; Johri, M.; Akaho, E.H.K.

2006-04-01

The Chinese Miniature Neutron Source Reactor (MNSR) is a low power research reactor with maximum thermal neutron flux of 1 x 10 12 n.cm -2 .s -1 in one of its inner irradiation channels and thermal power of approximately 30kW. The MNSR is designed based on the Canadian SLOWPOKE reactor and is one of the smallest commercial research reactors presently available in the world. Its commercial versions currently in operation in China, Ghana, Iran, Nigeria, Pakistan and Syria, is considered as an excellent tool for Neutron Activation Analysis (NAA), training of Scientist, and Engineers in nuclear science and technology and small scale radioisotope production. The paper highlights the basic design and theory of the commercial MNSR, its safety features, applications and advantages over the Chinese Prototype. The experimental flux characteristics determined in this work and in similar studies by other authors reveal that the commercial MNSR has more flux stability, longer life span, higher negative temperature coefficient of reactivity and low under-moderation compared to its prototype in China. The result shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application of NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. (author)

Prototype Learning and Dissociable Categorization Systems in Alzheimer’s Disease

Science.gov (United States)

Heindel, William C.; Festa, Elena K.; Ott, Brian R.; Landy, Kelly M.; Salmon, David P.

2015-01-01

Recent neuroimaging studies suggest that prototype learning may be mediated by at least two dissociable memory systems depending on the mode of acquisition, with A/Not-A prototype learning dependent upon a perceptual representation system located within posterior visual cortex and A/B prototype learning dependent upon a declarative memory system associated with medial temporal and frontal regions. The degree to which patients with Alzheimer’s disease (AD) can acquire new categorical information may therefore critically depend upon the mode of acquisition. The present study examined A/Not-A and A/B prototype learning in AD patients using procedures that allowed direct comparison of learning across tasks. Despite impaired explicit recall of category features in all tasks, patients showed differential patterns of category acquisition across tasks. First, AD patients demonstrated impaired prototype induction along with intact exemplar classification under incidental A/Not-A conditions, suggesting that the loss of functional connectivity within visual cortical areas disrupted the integration processes supporting prototype induction within the perceptual representation system. Second, AD patients demonstrated intact prototype induction but impaired exemplar classification during A/B learning under observational conditions, suggesting that this form of prototype learning is dependent on a declarative memory system that is disrupted in AD. Third, the surprisingly intact classification of both prototypes and exemplars during A/B learning under trial-and-error feedback conditions suggests that AD patients shifted control from their deficient declarative memory system to a feedback-dependent procedural memory system when training conditions allowed. Taken together, these findings serve to not only increase our understanding of category learning in AD, but to also provide new insights into the ways in which different memory systems interact to support the acquisition of

Irradiation of an uranium silicide prototype in RA-3 reactor; Irradiacion de un elemento combustible prototipo de siliciuro de uranio en el RA-3

Energy Technology Data Exchange (ETDEWEB)

Calabrese, R; Estrik, G; Notari, C [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

1997-12-31

The factibility of irradiation of an uranium silicide (U{sub 3} Si{sub 2}) prototype in the RA-3 reactor was studied. The standard RA-3 fuel element uses U{sub 3} O{sub 8} as fissible material. The enrichment of both standard and prototype is the same: 20% U{sub 235} and also the frame geometry and number of plates is identical. The differences are in the plate dimensions and the fissile content which is higher in the prototype. The cooling conditions of the core allow the insertion of the prototype in any core position, even near the water trap, if the overall power is kept below 5Mw. Nevertheless, the recommendation was to begin irradiation near the periphery and later on move the prototype towards more central positions in order to increase the burnup rate. The prototype was effectively introduced in a peripheral position and the thermal fluxes were measured between plates with the foil activation technique. These were also evaluated with the fuel management codes and a reasonable agreement was found. (author). 5 refs., 3 figs., 3 tabs.

High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

International Nuclear Information System (INIS)

Venkatraman, B.; Sethi, V.K.; Jayakumar, T.; Raj, B.

1995-01-01

In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

Gas leak tightness of SiC/SiC composites at elevated temperature

Energy Technology Data Exchange (ETDEWEB)

Hayasaka, Daisuke, E-mail: hayasaka@oasis.muroran-it.ac.jp [OASIS, Muroran Institute of Technology, Muroran, Hokkaido (Japan); Graduate School of Engineering, Muroran Institute of Technology, Muroran, Hokkaido (Japan); Park, Joon-Soo. [OASIS, Muroran Institute of Technology, Muroran, Hokkaido (Japan); Kishimoto, Hirotatsu [OASIS, Muroran Institute of Technology, Muroran, Hokkaido (Japan); Graduate School of Engineering, Muroran Institute of Technology, Muroran, Hokkaido (Japan); Kohyama, Akira [OASIS, Muroran Institute of Technology, Muroran, Hokkaido (Japan)

2016-11-01

Highlights: • NITE-SiC/SiC has extremely densified microstructure compared with other SiC/SiC composite like CVI. • Excellent helium and hydrogen gas-leak tightness of SiC/SiC composites by DEMO-NITE method from prototype industrialization production line was presented. • The excellence against stainless steel and Zircaloy at elevated temperature, together with generic excellent properties of SiC will be inevitable for innovative blanket and divertors for DEMO- and power- fusion reactors. - Abstract: SiC/SiC composite materials are attractive candidates for high heat flux components and blanket of fusion reactor, mainly due to their high temperature properties, radiation damage tolerance and low induced radioactivity. One of the challenges for SiC/SiC application in fusion reactors is to satisfy sufficient gas leak tightness of hydrogen and helium isotopes. Although many efforts have been carried-out, SiC/SiC composites by conventional processes have not been successful to satisfy the requirements, except SiC/SiC composites by NITE-methods. Toward the early realization of SiC/SiC components into fusion reactor systems process development of NITE-process has been continued. Followed to the brief introduction of recently developed DEMO-NITE process, baseline properties and hydrogen and helium gas leak tightness is presented. SiC/SiC claddings with 10 mm in diameter and 1 mm in wall thickness are tested by gas leak tightness system developed. The leak tightness measurements are done room temperature to 400 °C. Excellent gas leak tightness equivalent or superior to Zircaloy claddings for light water fission reactors is confirmed. The excellent gas leak tightness suggests nearly perfect suppression of large gas leak path in DEMO-NITE SiC/SiC.

Neutron radiography in the IEA-R1 reactor

International Nuclear Information System (INIS)

Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

1988-08-01

Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

SEMPaC - an expert system prototype associated with safety parameter display system of a nuclear power plant

International Nuclear Information System (INIS)

Hirama, K.

1989-01-01

This work presents SEMPaC, an expert system prototype: it provides means to support diagnosis and to make decisions during abnormal transients that cause the trip of nuclear power plant. The system operation is associated with Safety Parameter Display System - SPDS that was recommended by U. S. Nuclear Regulatory Commission (NRC) after the Three-Mile Island (TMI) accident analysis. (author)

Prototypical Consolidation Demonstration Project: Final report

International Nuclear Information System (INIS)

Gili, J.A.; Poston, V.K.

1993-11-01

This is the final report of the Prototypical Consolidation Demonstration Project, which was funded by the US Department of Energy's Office of Civilian Radioactive Waste Management. The project had two objectives: (a) to develop and demonstrate a prototype of production-scale equipment for the dry, horizontal consolidation and packaging of spent nuclear fuel rods from commercial boiling water reactor and pressurized water reactor fuel assemblies, and (b) to report the development and demonstration results to the US Department of Energy, Idaho Operations Office. This report summarizes the activities and conclusions of the project management contractor, EG ampersand G Idaho, Inc., and the fabrication and testing contractor, NUS Corporation (NUS). The report also presents EG ampersand G Idaho's assessments of the equipment and procedures developed by NUS

Analysis of gamma ray intensity on the S/C vent pipes area in the unit 2 reactor building of the Fukushima Daiichi Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Cho, Jai Wan; Choi, Young Soo; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The robot is equipped with cameras, a dosimeter, and 2 DOF (degree of freedom) manipulation arms. It loads a small vehicle equipped with a camera that can access and inspect narrow areas. TEPCO is using the four-legged walking robot to inspect the suppression chamber (S/C) area of the unit 2 reactor building basement in the Fukushima Daiichi Nuclear Power Plant. The robot carried out 6 missions for about four months, from 11 December, 2012 to 15 March, 2013, where it examined an evidence of a leakage of radioactivity contaminated water in the S/C area of unit 2 reactor building. When a camera's signal processing unit, which is consist of ASIC and FPGA devices manufactured by a CMOS fabrication process, is exposed to a higher dose rate gamma ray, the speckle distribution in the camera image increase more. From the inspection videos, released by TEPCO, of the underground 8 vent pipes in the unit 2 reactor building, we analyzed the speckle distribution from the high dose-rate gamma rays. Based on the distribution of the speckle, we attempted to characterize the vent pipe with much radioactivity contaminated materials among the eight vent pipes connected to the PCV. The numbers of speckles viewed in the image of a CCD (or CMOS) camera are related to an intensity of the gamma ray energy emitted by a nuclear fission reaction from radioactivity materials. The numbers of speckles generated by gamma ray irradiation in the camera image are calculated by an image processing technique. Therefore, calculating the speckles counts, we can determine the vent pipe with relatively most radioactivity-contaminated materials among the other vent pipes. From the comparison of speckles counts calculated in the inspection image of the vent pipe with the speckles counts extracted by gamma ray irradiation experiment of the same small vehicle camera model loaded with the four-legged walking robot, we can qualitatively estimate the gamma ray dose-rate in the S/C vent pipe area of the

OpenStreetMap Collaborative Prototype, Phase 1

Science.gov (United States)

Wolf, Eric B.; Matthews, Greg D.; McNinch, Kevin; Poore, Barbara S.

2011-01-01

Phase One of the OpenStreetMap Collaborative Prototype (OSMCP) attempts to determine if the open source software developed for the OpenStreetMap (OSM, http://www.openstreetmap.org) can be used for data contributions and improvements that meet or exceed the requirements for integration into The National Map (http://www.nationalmap.gov). OpenStreetMap Collaborative Prototype Phase One focused on road data aggregated at the state level by the Kansas Data Access and Support Center (DASC). Road data from the DASC were loaded into a system hosted by the U.S. Geological Survey (USGS) National Geospatial Technical Operations Center (NGTOC) in Rolla, Missouri. U.S. Geological Survey editing specifications were developed by NGTOC personnel (J. Walters and G. Matthews, USGS, unpub. report, 2010). Interstate and U.S. Highways in the dataset were edited to the specifications by NGTOC personnel while State roads were edited by DASC personnel. Resulting data were successfully improved to meet standards for The National Map once the system and specifications were in place. The OSM software proved effective in providing a usable platform for collaborative data editing

Prometheus Reactor I and C Software Development Methodology, for Action

International Nuclear Information System (INIS)

T. Hamilton

2005-01-01

The purpose of this letter is to submit the Reactor Instrumentation and Control (I and C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I and C Software Development Process Manual and Reactor Module Software Development Plan to NR for information

A Behavior-Preserving Translation From FBD Design to C Implementation for Reactor Protection System Software

International Nuclear Information System (INIS)

Yoo, Junbeom; Kim, Euisub; Lee, Jangsoo

2013-01-01

Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation

A Behavior-Preserving Translation From FBD Design to C Implementation for Reactor Protection System Software

Energy Technology Data Exchange (ETDEWEB)

Yoo, Junbeom; Kim, Euisub [Konkuk Univ., Seoul (Korea, Republic of); Lee, Jangsoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-08-15

Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation.

Macroscopic High-Temperature Structural Analysis Model of Small-Scale PCHE Prototype (II)

International Nuclear Information System (INIS)

Song, Kee Nam; Lee, Heong Yeon; Hong, Sung Deok; Park, Hong Yoon

2011-01-01

The IHX (intermediate heat exchanger) of a VHTR (very high-temperature reactor) is a core component that transfers the high heat generated by the VHTR at 950 .deg. C to a hydrogen production plant. Korea Atomic Energy Research Institute manufactured a small-scale prototype of a PCHE (printed circuit heat exchanger) that was being considered as a candidate for the IHX. In this study, as a part of high-temperature structural integrity evaluation of the small-scale PCHE prototype, we carried out high-temperature structural analysis modeling and macroscopic thermal and elastic structural analysis for the small-scale PCHE prototype under small-scale gas-loop test conditions. The modeling and analysis were performed as a precedent study prior to the performance test in the small-scale gas loop. The results obtained in this study will be compared with the test results for the small-scale PCHE. Moreover, these results will be used in the design of a medium-scale PCHE prototype

Prototyping Practice

DEFF Research Database (Denmark)

Ramsgaard Thomsen, Mette; Tamke, Martin

2015-01-01

This paper examines the role of the prototyping in digital architecture. During the past decade, a new research field has emerged exploring the digital technology’s impact on the way we think, design and build our environment. In this practice the prototype, the pavilion, installation or demonstr......This paper examines the role of the prototyping in digital architecture. During the past decade, a new research field has emerged exploring the digital technology’s impact on the way we think, design and build our environment. In this practice the prototype, the pavilion, installation...

Research reactor`s role in Korea

Energy Technology Data Exchange (ETDEWEB)

Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1996-12-31

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

Molten salt reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

2014-01-01

Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Sinha, N.K., E-mail: nksinha@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India); Raj, Baldev, E-mail: dir@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India)

2011-03-15

Research highlights: Production of thin fluoroelastomer profiles by cold feed extrusion and continuous cure involving microwave and hot air heating. Use of peroxide curing in air during production. Use of fluoroelastomers based on advanced polymer architecture (APA) for the production of profiles. Use of the profiles in inflatable seals for critical application of Prototype Fast Breeder Reactor. Tailoring of material formulation by synchronized optimization of material and production technologies to ensure that the produced seal ensures significant gains in terms of performance and safety in reactor under synergistic influences of temperature, radiation, air and sodium aerosol. - Abstract: The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of {approx}2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 {sup o}C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large

Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

International Nuclear Information System (INIS)

Sinha, N.K.; Raj, Baldev

2011-01-01

Research highlights: → Production of thin fluoroelastomer profiles by cold feed extrusion and continuous cure involving microwave and hot air heating. → Use of peroxide curing in air during production. → Use of fluoroelastomers based on advanced polymer architecture (APA) for the production of profiles. → Use of the profiles in inflatable seals for critical application of Prototype Fast Breeder Reactor. → Tailoring of material formulation by synchronized optimization of material and production technologies to ensure that the produced seal ensures significant gains in terms of performance and safety in reactor under synergistic influences of temperature, radiation, air and sodium aerosol. - Abstract: The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of ∼2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 o C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for

1/2-BPS correlators as c = 1 S-matrix

International Nuclear Information System (INIS)

Jevicki, Antal; Yoneya, Tamiaki

2007-01-01

We argue from two complementary viewpoints of Holography that the 2-point correlation functions of 1/2-BPS multi-trace operators in the large-N (planar) limit are nothing but the (Wick-rotated) S-matrix elements of c = 1 matrix model. On the bulk side, we consider an Euclideanized version of the so-called bubbling geometries and show that the corresponding droplets reach the conformal boundary. Then the scattering matrix of fluctuations of the droplets gives directly the two-point correlators through the GKPW prescription. On the Yang-Mills side, we show that the two-point correlators of holomorphic and anti-holomorphic operators are essentially equivalent with the transformation functions between asymptotic in- and out-states of c = 1 matrix model. Extension to non-planar case is also discussed

Prototypical spent fuel rod consolidation equipment preliminary design report: Volume 1, Report

International Nuclear Information System (INIS)

1986-01-01

This design report describes the NUS Preliminary Design of the Prototype Spent Nuclear Fuel Rod Consolidation Equipment for the Department of Energy. The sections of the report elaborate on each facet of the preliminary design. A concept summary is provided to assist the reader in rapidly understanding the complete design. The NUS Prototype Spent Fuel Rod Consolidation System is an automatically controlled system to consolidate a minimum of 750 MT (heavy metal)/year of US commercial nuclear reactor fuel, at 75% availability. The system is designed with replaceable components utilizing the latest state-of-the-art technology. This approach gives the system the flexibility to be developed without costly development programs, yet accept new technology as it evolves over the next ten years. Capability is also provided in the system design to accommodate a wide variety of fuel conditions and to recover from any situation which may arise

International breeder reactor development

International Nuclear Information System (INIS)

Traube, K.

1976-01-01

For more than a decade, sodium cooled breeder reactors have now been in the focus of advanced nuclear power development in the major industrialized countries. In the sixties, a total of seven small experimental nuclear power stations were commissioned. Two of these have been shut down in the meantime, the others continue to work satisfactorily, their main purpose being the development of fuel elements. The years 1972-1974 saw the commissioning of the prototype power stations in the 300 MWe power category in France, the United Kingdom and the Soviet Union. Presently, other experimental reactors are under construction in the Federal Republic of Germany, Italy, Japan, the United States, plus another Soviet 600 MWe prototype reactor and the SNR 300 DeBeNeLux prototype at Kalkar. A comparison of the technological features either implemented or planned in the prototype and experimental power plants and of their fuel elements reveals a remarkable similarity in the basic concepts pursued in different countries. The two types of breeder reactors, viz. the loop and the pool types, show a closer resemblance to each other than do pressurized and boilling water reactors. The growing awareness of administrative problems emerging in the approaching phase of the introduction of large breeder power stations in a number of European countries has recently led to a streamlining effort in the structure of industries and to tentative steps towards international cooperation on a broad basis. (orig.) [de

Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

Energy Technology Data Exchange (ETDEWEB)

M. J. Appel and J. M. Capron

2007-07-25

This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

International Nuclear Information System (INIS)

Appel, M.J.; Capron, J.M.

2007-01-01

This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes

Rietveld refinement of the structures of 1.0 C-S-H and 1.5 C-S-H

KAUST Repository

Battocchio, Francesco

2012-11-01

Low-Q region Rietveld analyses were performed on C-S-H synchrotron XRD patterns, using the software MAUD. Two different crystal structures of tobermorite 11 Å were used as a starting model: monoclinic ordered Merlino tobermorite, and orthorhombic disordered Hamid tobermorite. Structural modifications were required to adapt the structures to the chemical composition and the different interlayer spacing of the C-S-H samples. Refinement of atomic positions was done by using special constraints called fragments that maintain interatomic distances and orientations within atomic polyhedra. Anisotropic crystallite size refinement showed that C-S-H has a nanocrystalline disordered structure with a preferred direction of elongation of the nanocrystallites in the plane of the Ca interlayer. The quality of the fit showed that the monoclinic structure gives a more adequate representation of C-S-H, whereas the disordered orthorhombic structure can be considered a more realistic model if the lack of long-range order of the silica chain along the c-direction is assumed. © 2012 Elsevier Ltd. All rights reserved.

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

Energy Technology Data Exchange (ETDEWEB)

GERBER MS

2009-04-28

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

Biodegradation of a commercial mixture of the herbicides atrazine and S-metolachlor in a multi-channel packed biofilm reactor.

Science.gov (United States)

Cabrera-Orozco, Alberto; Galíndez-Nájera, Silvia Patricia; Ruiz-Ordaz, Nora; Galíndez-Mayer, Juvencio; Martínez-Jerónimo, Fernando

2017-11-01

Atrazine and S-metolachlor are two of the most widely used herbicides for agricultural purposes; consequently, residues of both compounds and their metabolites had been detected in ground and superficial waters. Unlike atrazine, the complete degradation of metolachlor has not been achieved. Hence, the purpose of this research is to study the biodegradation of a commercial mixture of atrazine and S-metolachlor in a prototype of a multi-channel packed-bed-biofilm reactor (MC-PBR) designed with the aim of solving the problems of pressure drop and oxygen transfer, typically found on this type of bioreactors.Because the removal efficiency of the herbicides was increased when Candida tropicalis was added to the original microbial community isolated, the reactor was inoculated with this enriched community. The operational conditions tested in batch and continuous mode did not affect the removal efficiency of atrazine; however, this was not the case for S-metolachlor. The removal rates and efficiencies showed a notable variation along the MC-PBR operation.

Risk-oriented analysis of the German prototype fast breeder reactor SNR-300: off-site accident consequence model and results of the study

International Nuclear Information System (INIS)

Bayer, A.; Ehrhardt, J.

1984-01-01

Accident off-site consequence calculations and risk assessments performed for the ''risk oriented analysis'' of the German prototype fast breeder reactor SNR-300 were performed with a modified version of the off-site accident consequence model UFOMOD. The modifications mainly relate to the deposition and resuspension processes, the ingestion model, and the dose factors. Consequence calculations at the site of Kalkar on the Rhine River were performed for 115 weather sequences in 36 wind directions. They were based on seven release categories evaluated for the SNR-300 with two different fueling strategies: plutonium from Magnox reactors only and plutonium from light water reactors and Magnox reactors. In parallel, the corresponding frequencies of occurrence are determined. The following results are generated: 1. complementary cumulative frequency distribution functions for collective fatalities and collective doses 2. expected values of the collective fatalities and collective doses as well as distance-dependent expected values of individual fatality 3. contributions of the different exposure pathways to fatalities with respect to the various organs. For comparison with the risk of a PWR-1300, calculations for the PWR-1300 of the ''German Risk Study'' were repeated with the same modified consequence model. Comparison shows that smaller risks result for the SNR-300. However, the confidence interval bandwidths obtained for the frequencies of the release categories for the SNR-300 are larger than those of the PWR-1300

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Irradiation routine in the IPR-R1 Triga reactor

International Nuclear Information System (INIS)

Maretti Junior, F.

1980-01-01

Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

General Aspects of CAREM-25 Reactor

International Nuclear Information System (INIS)

Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

2000-01-01

CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

Erratum!- in paper volume 19, Supplement 1, Year 2015, pp. S109-S115, DOI:10.2298/TSCI15S1S09C

Directory of Open Access Journals (Sweden)

Editorial

2015-01-01

Full Text Available In the paper published in THERMAL SCIENCE Volume 19, Supplement 1, YEAR 2015, pp. S109-S115, DOI REFERENCE: 10.2298/TSCI15S1S09C Names and affiliations of the authors has been incorrectly written Istead of: THE DIFFUSION MODEL OF FRACTAL HEAT AND MASS TRANSFER IN FLUIDIZED BED A Local Fractional Arbitrary Euler-Lagrange Formula by Xu CHENG∗ and Xiao-Xun MA School of Chemical Engineering, Northwest University, Xi'an, Shaanxi, China Correctly has to be written: THE DIFFUSION MODEL OF FRACTAL HEAT AND MASS TRANSFER IN FLUIDIZED BED A Local Fractional Arbitrary Euler-Lagrange Formula by Xu CHENG1∗ and Lin Wang2 - 1School of Chemical Engineering, Northwest University, Xi'an, Shaanxi, China - 2Xi'an Modern Chemistry Research Institute, Xi'an, Shaanxi, China Link to the corrected article 10.2298/TSCI15S1S09C

39 CFR 211.1 - Disposition of former title 39, U.S.C.

Science.gov (United States)

2010-07-01

... 39 Postal Service 1 2010-07-01 2010-07-01 false Disposition of former title 39, U.S.C. 211.1... REGULATIONS § 211.1 Disposition of former title 39, U.S.C. Except as otherwise continued in effect as postal regulations, all provisions of former title 39, U.S.C., which were continued in effect as regulations of the...

Pathogenesis comparison between the United States porcine epidemic diarrhoea virus prototype and S-INDEL-variant strains in conventional neonatal piglets.

Science.gov (United States)

Chen, Qi; Gauger, Phillip C; Stafne, Molly R; Thomas, Joseph T; Madson, Darin M; Huang, Haiyan; Zheng, Ying; Li, Ganwu; Zhang, Jianqiang

2016-05-01

At least two genetically different porcine epidemic diarrhoea virus (PEDV) strains have been identified in the USA: US PEDV prototype and S-INDEL-variant strains. The objective of this study was to compare the pathogenicity differences of the US PEDV prototype and S-INDEL-variant strains in conventional neonatal piglets under experimental infections. Fifty PEDV-negative 5-day-old pigs were divided into five groups of ten pigs each and were inoculated orogastrically with three US PEDV prototype isolates (IN19338/2013, NC35140/2013 and NC49469/2013), an S-INDEL-variant isolate (IL20697/2014), and virus-negative culture medium, respectively, with virus titres of 104 TCID50 ml- 1, 10 ml per pig. All three PEDV prototype isolates tested in this study, regardless of their phylogenetic clades, had similar pathogenicity and caused severe enteric disease in 5-day-old pigs as evidenced by clinical signs, faecal virus shedding, and gross and histopathological lesions. Compared with pigs inoculated with the three US PEDV prototype isolates, pigs inoculated with the S-INDEL-variant isolate had significantly diminished clinical signs, virus shedding in faeces, gross lesions in small intestines, caeca and colons, histopathological lesions in small intestines, and immunohistochemistry staining in ileum. However, the US PEDV prototype and the S-INDEL-variant strains induced similar viraemia levels in inoculated pigs. Whole genome sequences of the PEDV prototype and S-INDEL-variant strains were determined, but the molecular basis of virulence differences between these PEDV strains remains to be elucidated using a reverse genetics approach.

Some general remarks concerning the experimental devices and means for flux measurements used in the research reactors of the C.E.A

International Nuclear Information System (INIS)

Rossillon, F.; Chauvez, C.

1964-01-01

First, the authors define the operation and organisation lay-out of the research reactors. This out lay is then used as a basis for an examination of the conception, realisation and exploitation of irradiation devices. They then analyse the different operations carried out in placing an irradiation device in the reactor. They point out that the time necessary for decision of construction, conception, for material provision, for construction, out-of-pile tests and security examination is in general longer than the irradiation time. To minimize this time they advise a closer collaboration between the personnel of the experimentation laboratory and the reactor personnel for the conception of the design. They think it would be preferable to construct the prototypes in the research centre and leave the fabrication of small series to the industry. The methods used for the in-pile operation of the device are also indicated. Considering the evolution of the construction techniques they mention that the requirements for the irradiation conditions are put higher and higher (reactor stability and homogeneity). Also described are the performances of the main irradiation devices in operation in research reactor in the C.E.A. (authors) [fr

Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

International Nuclear Information System (INIS)

Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

2016-01-01

Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

Fast breeder reactors

International Nuclear Information System (INIS)

Heinzel, V.

1975-01-01

The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

Potentiation of C1-esterase inhibitor by heparin and interactions with C1s protease as assessed by surface plasmon resonance.

Science.gov (United States)

Rajabi, Mohsen; Struble, Evi; Zhou, Zhaohua; Karnaukhova, Elena

2012-01-01

Human C1-esterase inhibitor (C1-INH) is a multifunctional plasma protein with a wide range of inhibitory and non-inhibitory properties, mainly recognized as a key down-regulator of the complement and contact cascades. The potentiation of C1-INH by heparin and other glycosaminoglycans (GAGs) regulates a broad spectrum of C1-INH activities in vivo both in normal and disease states. SCOPE OF RESEARCH: We have studied the potentiation of human C1-INH by heparin using Surface Plasmon Resonance (SPR), circular dichroism (CD) and a functional assay. To advance a SPR for multiple-unit interaction studies of C1-INH we have developed a novel (consecutive double capture) approach exploring different immobilization and layout. Our SPR experiments conducted in three different design versions showed marked acceleration in C1-INH interactions with complement protease C1s as a result of potentiation of C1-INH by heparin (from 5- to 11-fold increase of the association rate). Far-UV CD studies suggested that heparin binding did not alter C1-INH secondary structure. Functional assay using chromogenic substrate confirmed that heparin does not affect the amidolytic activity of C1s, but does accelerate its consumption due to C1-INH potentiation. This is the first report that directly demonstrates a significant acceleration of the C1-INH interactions with C1s due to heparin by using a consecutive double capture SPR approach. The results of this study may be useful for further C-INH therapeutic development, ultimately for the enhancement of current C1-INH replacement therapies. Published by Elsevier B.V.

SIMS prototype system 1 test results: Engineering analysis

Science.gov (United States)

1978-01-01

The space and domestic water solar heating system designated SIMS Prototype Systems 1 was evaluated. The test system used 720 ft (gross) of Solar Energy Products Air Collectors, a Solar Control Corporation SAM 20 Air Handler with Model 75-175 control unit, a Jackson Solar Storage tank with Rho Sigma Mod 106 controller, and 20 tons of rack storage. The test data analysis performed evaluates the system performance and documents the suitability of SIMS Prototype System 1 hardware for field installation.

Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

2007-09-01

The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

Representations of Nets of C*-Algebras over S 1

Science.gov (United States)

Ruzzi, Giuseppe; Vasselli, Ezio

2012-11-01

In recent times a new kind of representations has been used to describe superselection sectors of the observable net over a curved spacetime, taking into account the effects of the fundamental group of the spacetime. Using this notion of representation, we prove that any net of C*-algebras over S 1 admits faithful representations, and when the net is covariant under Diff( S 1), it admits representations covariant under any amenable subgroup of Diff( S 1).

Summary report on safety and licensing strategy support for the ABR prototype

International Nuclear Information System (INIS)

Cahalan, J. E.; Nuclear Engineering Division

2007-01-01

Argonne National Laboratory is providing support to the US Department of Energy in the Global Nuclear Energy Partnership (GNEP) in certification of an advanced, sodium-cooled fast reactor. The reactor is to be constructed as a prototype for future commercial power reactors that will produce electricity while consuming actinides recovered from light water reactor spent fuel. This prototype reactor has been called the Advanced Burner Reactor, or ABR, and is now often referred to as the advanced recycle reactor. As part of its activities, Argonne is providing technical services to assist definition of a safety and licensing strategy for the ABR prototype, and to further implementation of the strategy. In FY06, an organizational meeting was held for DOE and its laboratory contractors to discuss licensing alternatives and review previous licensing experience for the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor Plant (CRBRP). Near the end of FY06, a report summarizing the discussions and conclusions was written. One of the top-level conclusions recorded in the report was a recommendation to follow a licensing strategy that included the US Nuclear Regulatory Commission (NRC) as the regulatory review and licensing authority. In FY07, activities ar Argonne to support safety and licensing progress have continued. These activities have focused on further evaluation of licensing alternatives; assessment of design, analysis, and documentation implications of licensing paths; and initial technical interactions with the Nuclear Regulatory Commission. This report summarizes FY07 activities

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Occupied and unoccupied orbitals of C{sub 60} and C{sub 70} probed with C 1s emission and absorption

Energy Technology Data Exchange (ETDEWEB)

Carlisle, J.A.; Terminello, L.J.; Hudson, E.A. [Lawrence Berkeley National Lab., CA (United States)] [and others

1997-04-01

The aim of this work is to characterize the orbital structure of the fullerenes, and to pursue its evolution from a cluster to the infinite solid. For obtaining a complete picture of the electronic structure the authors compare a variety of experimental techniques, i.e. photoemission and core level emission for occupied orbitals and inverse photoemission and core level absorption for unoccupied orbitals. Their experimental results focus on optical probes involving the C 1s core level, i.e. absorption via transitions from the C 1s level into unoccupied {pi}* and {sigma}* orbitals and emission involving transitions from occupied orbitals into a C 1s hole. Due to the simplicity of the C 1s level there exist clear selection rules. For example, only transitions to and from orbitals with p-character are dipole-allowed. These results on the p-projected density of states are compared with inverse photoemission and photoemission results, where the selection rules are less definitive. In addition, a first-principles quasiparticle calculation of the density of states is used to assign the orbital features. The spectra from C{sub 60} and C{sub 70} are still far from their infinite analog, i.e., graphite, which is also measured with the same techniques. In order to determine the effect of electron transfer onto C{sub 60}, as in superconducting alkali fullerides, the authors are studying resonant emission of C{sub 60}. An electron is placed in the lowest unoccupied molecular orbital (LUMO) by optical absorption from the C 1s level and the C 1s emission detected in the presence of this spectator electron.

Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

International Nuclear Information System (INIS)

Kawahara, Toshio; Matsushita, Tadashi

1977-01-01

The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

Formation and early hydration characteristics of C2.75B1.25A3$ in binary system of C2.75B1.25A3$-C2S

Directory of Open Access Journals (Sweden)

Wang, Shoude

2016-09-01

Full Text Available C2.75B1.25A3$ (2.75CaO•1.25BaO• 3Al2O3• SO3 is one of the important minerals and it govern-directly the early-strength of belite-barium calcium sulphoaluminate cement. In this paper a binary system C2.75B1.25A3$-C2S is selected to investigate the formation of C2.75B1.25A3$. In the range of 1100 °C–1200 °C, the earlier formed C2S hinders the formation of C2.75B1.25A3$. On the contrary, when the temperature is in the range of 1200 °C–1350 °C, the initially formed C2S could provide a surface for the nucleation of C2.75B1.25A3$ and cut down the potential barrier (?Gk* for the heterogeneous nucleation of C2.75B1.25A3$, which contributes to its formation. Moreover, at 1350 °C, the large amount of previously formed C2S benefits the extent of formation of C2.75B1.25A3$. The possible reason was that it could prevent sulfur evaporation. In early hydration age, AFm and AFt originating from C2.75B1.25A3$ hydration are found within 2 h and 12 h under 95% RH at 1 °C, respectively, whereas C2S is unhydrated at this moment.En el cemento de sulfoaluminato de calcio y bario, el C2.75B1.25A3$ (2.75CaO•1.25BaO• 3Al2 O3• SO3 es una de las principales fases, y regula directamente la resistencia inicial del cemento. En este trabajo, se ha seleccionado el sistema binario C2.75B1.25A3$-C2S para investigar la formación de C2.75B1.25A3$. En el rango de 1100 °C-1200 °C, el C2S formado anteriormente impide la formación de C2.75B1.25A3$, mientras que cuando la temperatura está entre 1200 °C-1350 °C, el C2S proporcionaría una superficie de nucleación de C2.75B1.25A3$ reduciendo la barrera de potencial (?Gk* para la nucleación heterogénea de C2.75B1.25A3$, lo que contribuye a su formación. Además, a 1350 °C, la gran cantidad de C2S formado beneficia la formación de C2.75B1.25A3$, ya que podía prevenir la evaporación del azufre. En las primeras etapas de la hidratación (entre 2 y 12h y 95% HR a 1 ºC se pueden encontrar AFM y AFt

Reactor operations at SAFARI-1

International Nuclear Information System (INIS)

Vlok, J.W.H.

2003-01-01

A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

U.S. Nuclear Power Reactor Plant Status

Data.gov (United States)

Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

Status and program of development of the fast breeder reactor system in the U.S

International Nuclear Information System (INIS)

Roberts, R.W.

1977-01-01

The U.S. Energy Research and Development Administration's highest priority breeder concept is the Liquid Metal Fast Breeder Reactor (LMFBR). Its objective is the development of a broad technological and engineering base with extensive utility and industrial involvement which will establish a timely capability for a competitive and commercial breeder industry. In addition to technological development, an integral part of the LMFBR program is the transfer of LMFBR technology to the nuclear industry and the parallel development of utility capability. An important indicator of the effectiveness of this program is the successful construction and operation of a number of LMFBR experimental and test reactors and associated non-nuclear test facilities such as those located at Santa Susana, California, the Argonne (III). National Laboratory and the Hanford Engineering Development Laboratory at Richland, Washington. A principal element of the U.S. fast breeder reactor program is the Fast Flux Test Facility (FFTF). The largest test reactor facility in the world, the FFTF will focus on the testing of fuels and material. The FFTF, being built at the Hanford Engineering Development Laboratory, is scheduled for completion in the late 1970's. The next step toward timely commercial breeder capability is to implement a program for a large-scale demonstration of this concept. A 380 megawatt electrical demonstration plant - designated the Clinch River Breeder Reactor Plant (CRBRP) - will be constructed and operated near Oak Ridge, Tennessee, under a cooperative arrangement of industrial contractors, utilities, and the Government. The completion date is 1983. The estimated cost of the CRBRP project is 1.95 billion dollars. This includes the cost of design, construction, related research and development, and five years of operation - the full timespan from the project's beginning in 1972 through completion. The Nation's principal electric utilities have strongly endorsed the

Pulsed Nd-YAG laser welding of Prototype Fast Breeder Reactor fuel elements

International Nuclear Information System (INIS)

Suresh Varma, P.V.; Gupta, Amit; Amit, K.; Bhatt, R.B.; Afzal, Mohd.; Panakkal, J.P.; Kamath, H.S.

2009-02-01

End plug welding of Prototype Fast Breeder Reactor (PFBR) fuel elements involves welding of fully Austenitic Stainless Steel (ASS) of grade D9 clad tube with 316M end plug. Pulsed Gas Tungsten Arc Welding (GTAW) is being used for the production of PFBR fuel elements at Advanced Fuel Fabrication Facility (AFFF). GTAW is an established process for end plug welding and hence adopted by many countries. GTAW has got certain limitations like heat input, arc gap sensitivity and certain sporadic defects like tungsten inclusion. Experiments have been carried out at AFFF to use Laser Beam Welding (LBW) technique as LBW offers a number of advantages over the former process. This report mainly deals with the optimization of laser parameters for welding of PFBR fuel elements. To facilitate pulsed Nd-YAG laser spot welding, parameters like peak power, pulse duration, pulse energy, frequency and defocusing of laser beam on to the work piece have been optimized. On the basis of penetration requirement laser welding parameters have been optimized. (author)

Performance of a UA1 hadron calorimeter prototype

International Nuclear Information System (INIS)

Corden, M.J.; Dowell, J.D.; Edwards, M.; Ellis, N.; Garvey, J.; Grant, D.; Homer, R.J.; Kenyon, I.R.; McMahon, T.; Schanz, G.; Sumorok, K.C.T.O.; Watkins, P.M.; Wilson, J.A.; Eisenhandler, E.; Gibson, W.R.; Kalmus, P.I.P.; Thompson, G.; Arnison, G.; Astbury, A.; Grayer, G.; Haynes, W.J.; Hill, D.; Nandi, A.K.; Roberts, C.; Shah, T.P.

1982-01-01

The hadron calorimeter for the UA1 experiment at the CERN SPS proton-antiproton collider consists of a lead-scintillator sandwich plus an iron-scintillator sandwich with wavelength shifter readout. The authors have tested prototype modules in muon and hadron beams in the momentum range from 0.7 to 90 GeV/c. For several angles of incidence, the authors have studied the uniformity of the response to hadrons as a function of position. This has included regions where there is reduced sensitivity due to mechanical constraints and the presence of the wavelength shifter readout. The response, resolution and degree of shower containment were measured as a function of incident momentum. (Auth.)

C-Reactor I and E loading instability limits

Energy Technology Data Exchange (ETDEWEB)

Hess, K.W.

1957-01-24

The pilot charging of I & E fuel elements has been implemented at C-Reactor under Production Test IP-19-A. It was necessary to provide adequate tube protection against flow interruption by establishing proper trip setting on the Panellit pressure gauges. the administration of these Panellit trip settings is done by trip-before- boiling tube outlet temperature limits, which are similar in principle to the current instability limits. Trip-before-boiling limits for C-Reactor I & E fuel elements loadings are presented in this document.

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

AREVA's nuclear reactors portfolio

International Nuclear Information System (INIS)

Marincic, A.

2009-01-01

A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

Gender Prototype Representation in Media: Case Study of Hamshahri’s Events Page

Directory of Open Access Journals (Sweden)

Heshmat Sadat Moini Far

2009-11-01

Full Text Available Besides family and education system, media also affects on the socializing process. What media represent is mutually affected and affective with culture. However, there are some times when media’s effect on society is especially important. For example, people with restricted social interactions would spend more time watching TV; thus what they watch not only entertains, but also has educational effects on them, this is the same for journals with habitual readers. Reviewing Hamshahri’s events page, this study attempts to study social construction of gender prototypes. Applying gender role theory and theory of gender sociability, this study would offer a hybrid model. As results show, men are more inclined to commit a crime, and women are usually the object of crimes. The most sensible prototype is that men are represented as “aggressive and raper”, while women are the victims. Described prototypes are both descriptive and normative.

Generation IV reactors: economics

International Nuclear Information System (INIS)

Dupraz, B.; Bertel, E.

2003-01-01

The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

General aspects of CAREM-25 reactor

International Nuclear Information System (INIS)

Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

2000-01-01

CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

Prototype studies of a 1 MHz chopper for the KAON Factory

International Nuclear Information System (INIS)

Wait, G.D.; Barnes, M.J.; Bishop, D.; Waters, G.; Figley, C.B.

1991-05-01

A 1.025 MHz (≅ 10 6 discrete pulses/s) beam chopper is required for the injection line into the accumulator ring of the KAON Factory at TRIUMF. The beam chopper will create 108 ns gaps in the 1 GeV/c H - beam to allow enough time for the magnetic field to be established in the kicker magnets in each of the 5 rings. The required deflection of 1 mrad can be achieved with a set of plates 5 cm apart in which the product of voltage difference and plate length is 37.7 kV.m. The 'kick' must have a rise and fall time of less than 39 ns and a flat top of 49 ns on alternate pulses. A novel design concept for a 1 MHz chopper has been developed involving an energy storage system where the electric pulses are stored in a large diameter (10 cm) low loss coaxial cable. Measurements on the performance of a high voltage prototype are presented. Results are encouraging and show that this novel design can be implemented successfully for the KAON Factory. (Author) 11 refs., 6 figs., tab

Courthouse Prototype Building

Energy Technology Data Exchange (ETDEWEB)

Malhotra, Mini [ORNL; New, Joshua Ryan [ORNL; Im, Piljae [ORNL

2018-02-01

As part of DOE's support of ANSI/ASHRAE/IES Standard 90.1 and IECC, researchers at Pacific Northwest National Laboratory (PNNL) apply a suite of prototype buildings covering 80% of the commercial building floor area in the U.S. for new construction. Efforts have started on expanding the prototype building suite to cover 90% of the commercial building floor area in the U.S., by developing prototype models for additional building types including place of worship, public order and safety, public assembly. Courthouse is courthouse is a sub-category under the “Public Order and Safety" building type category; other sub-categories include police station, fire station, and jail, reformatory or penitentiary.ORNL used building design guides, databases, and documented courthouse projects, supplemented by personal communication with courthouse facility planning and design experts, to systematically conduct research on the courthouse building and system characteristics. This report documents the research conducted for the courthouse building type and proposes building and system characteristics for developing a prototype building energy model to be included in the Commercial Building Prototype Model suite. According to the 2012 CBECS, courthouses occupy a total of 436 million sqft of floor space or 0.5% of the total floor space in all commercial buildings in the US, next to fast food (0.35%), grocery store or food market (0.88%), and restaurant or cafeteria (1.2%) building types currently included in the Commercial Prototype Building Model suite. Considering aggregated average, courthouse falls among the larger with a mean floor area of 69,400 sqft smaller fuel consumption intensity building types and an average of 94.7 kBtu/sqft compared to 77.8 kBtu/sqft for office and 80 kBtu/sqft for all commercial buildings.Courthouses range in size from 1000 sqft to over a million square foot building gross square feet and 1 courtroom to over 100 courtrooms. Small courthouses

Gas Cooled Fast Reactors: Recent advances and prospects

International Nuclear Information System (INIS)

Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

2013-01-01

Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

Engineering Design of a Double Reactor for Spent Fuel Oxidation

International Nuclear Information System (INIS)

Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee

2015-01-01

In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more

Engineering Design of a Double Reactor for Spent Fuel Oxidation

Energy Technology Data Exchange (ETDEWEB)

Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more.

Material choices for the commercial fast reactor steam generators

International Nuclear Information System (INIS)

Willby, C.; Walters, J.

1978-01-01

Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

In-reactor fatigue crack propagation

International Nuclear Information System (INIS)

Ermi, A.M.; Mervyn, D.A.; Straalsund, J.L.

1979-08-01

An in-reactor fatigue experiment is being designed to determine the effect of dynamic irradiation on the fatigue crack propagation (FCP) behavior of candidate fusion first wall materials. This investigation has been prompted by studies which show gross differences in crack growth characteristics of creep rupture specimens testing by postirradiation versus dynamic in-reactor methods. The experiment utilizes miniature center-cracked-tension specimens developed specifically for in-reactor studies. In the test, a chain of eight specimens, precracked to various initial crack lengths, is stressed during irradiation to determine crack growth rate as a function of stress intensity. Load levels were chosen which result in small crack growth rates encompassing a regime of the crack growth curve not previously investigated during irradiation studies of FCP. The test will be conducted on 20% cold worked 316 stainless steel at a temperature of 425 0 C, in a sodium environment, and at a frequency of 1 cycle/min. Irradiation will occur in the Oak Ridge Research Reactor, resulting in a He/dpa ratio similar to that expected at the first wall in a fusion reactor. Detailed design of the experiment is presented, along with crack growth data obtained from prototypic testing of the experimental apparatus. These results are compared to data obtained under similar conditions generated by conventional test methods

S.A.C.I.: Incident Combat Support System; S.A.C.I.: Sistema de Apoio ao Combate de Incidentes

Energy Technology Data Exchange (ETDEWEB)

Macedo, Antonio R.L. [ARMTEC Tecnologia em Robotica, Fotrtaleza, CE (Brazil); Macedo, Antonio R.M. [Universidade de Fortaleza, CE (Brazil)

2004-07-01

The incidents that occur in the petrochemical industry are extremely dangerous, because of the range of temperature that it reaches and the radius of the explosion. For this reason the S.A.C.I. that is an Incident's Combat Support System was developed. The purpose of this paper is to present the complete operational capability of this machine, and also some of the construction design calculations. It is a controlled-by-distance robot that carries one water cannon that generates fog, stream or foam with a limit pressure of 125 psi. It works within 90 m from the operator, has 3 degrees of freedom and a minimum autonomy of 3 hours. Before this prototype was made, only the United Kingdom by Qinetiq and the Japan by the Tokyo Fire Department had this technology. This prototype is around 70% of the investment of the ones in the market. The tests shown in the paper were made in the training bunker of Ceara's Military Fire Corp. Headquarters and in an arena in the Gloria Marine in Rio de Janeiro. The results of this project is a national product that improves the incident's combat response time, saving the most important resource, that is the human been. (author)

An abnormal event advisory expert system prototype for reactor operators

International Nuclear Information System (INIS)

Hance, D.C.

1989-01-01

Nuclear plant operators must respond correctly during abnormal conditions in the presence of dynamic and potentially overwhelming volumes of information. For this reason, considerable effort has been directed toward the development of nuclear plant operator aids using artificial intelligence techniques. The objective of such systems is to diagnose abnormal conditions within the plant, possibly predict consequences, and advise the operators of corrective actions in a timely manner. The objective of the work is the development of a prototype expert system to diagnose abnormal events at a nuclear power plant and advise plant operators of the event and applicable procedures in an on-line mode. The major difference between this effort and previous work is the use of plant operating procedures as a knowledge source and as an integral part of the advice provided by the expert system. The acceptance by utilities of expert systems as operator aids requires that such systems be compatible with the regulatory environment and provide economic benefits. For this reason, commercially viable operator aid systems developed in the near future must complement existing plant procedures rather than reach beyond them in a revolutionary manner. A knowledge source is the resource providing facts and relationships that are coded into the expert system program. In this case, the primary source of knowledge is a set of selected abnormal operating procedures for a modern Westinghouse pressurized water reactor

Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

1994-01-01

The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

Synthesis of racemic, R- and S-[1-11C]-β-hydroxybutyric acid

International Nuclear Information System (INIS)

Thorell, J.-O.; Stone-Elander, S.; Karolinska Hospital and Inst., Stockholm; Koenig, W.A.; Halldin, C.; Widen, L.

1991-01-01

Racemic, R- and S-β-hydroxybutyric acid were labelled with 11 C in the carboxylic position by a two-step stereospecific synthesis starting with carrier-added [ 11 C]cyanide and R/S, R- or S-propylene oxide. Hydrolysis of the intermediate nitrile with hydrochloric acid gave racemic [1- 11 C]-β-hydroxybutyric acid and R- or S-[1- 11 C]-β-hydroxybutyric acid with an enantiomeric excess of 87-97%. The total synthesis time (including HPLC purification) was 45-50 min from end of trapping. The isolated decay-corrected radiochemical yield was 20-30% based on [ 11 C]cyanide. The radiochemical purity of the products was > 99%]. (author)

Summary of the Minor Actinide-bearing MOX AFC-2C and -2D Irradiations

International Nuclear Information System (INIS)

McClellan, Kenneth; Chichester, Heather; Hayes, Steve; Voit, Stewart

2013-01-01

Summary of AFC-2C and AFC-2D tests: • AFC-2C and 2D, 1st MOX experiments in FCRD, were irradiated in ATR; • Initial results indicate performance of experimental MA-MOX fuels are similar to standard FR MOX fuels; • Cd-shrouded ATR experiment assembly and 235 U enrichment produce prototypic fast reactor power and temperature profiles leading to classic MOX zone restructuring; • Baseline postirradiation examinations have been completed for AFC-2C MOX and MA-MOX fuels; • Future work includes: – PIE of AFC-2D; – compare results to prototypic MOX fuel performance; – electron microscopy for microstructure and constituent distribution; – advanced NDE on saved pins

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

Energy Technology Data Exchange (ETDEWEB)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Synthesis of (R,S)-[2,3-13C2]-1-(1'-methyl-2'-pyrrolidinyl)propan-2-one; {(R,S)-[2',3'-13C2]hygrinePound right bracePound

International Nuclear Information System (INIS)

Abraham, T.W.; Leete, Edward

1996-01-01

2-Ethoxy-1-methyl-5-pyrrolidinone (1) was reacted with ethyl [3,4- 13 C 2 ]-acetoacetate (2) in the presence of TiCl 4 to give ethyl [3,4- 13 C 2 ]-2-(1'-methyl-5'-oxo-2'-pyrrolidinyl)-3-oxobutanoate (3) in 85% yield. Decarboethoxylation of ethyl [3,4- 13 C 2 ]-2-(1'-methyl-5'-oxo-2'-pyrrolidinyl)-3-oxobutan-oate (3) was accomplished using NaCl and H 2 O in DMSO to give (R,S)-[2,3- 13 C 2 ]-1-(1'-methyl-5'-oxo-2'-pyrrolidinyl)propan-2-o ne (4) in 91% yield. Protection of the ketone as a ketal (ethylene glycol, H + ), followed by reduction of the amide to the amine using LiAlH 4 and subsequent deprotection of the ketal gave (R,S)-[2,3- 13 C 2 ]-1-(1'-methyl-2'-pyrrolidinyl)propan-2-one ((R,s)-[2', 3'- 13 C 2 ]Hygrine) (8) in 78% yield. (61% overall yield from ethyl [3,4- 13 C 2 ]acetoacetate). (Author)

Tasks for development of SiC/SiC composites as structural material in a fusion reactor

International Nuclear Information System (INIS)

Yamada, Reiji

1997-01-01

SiC/SiC composites are chosen for a structural material of blankets in DREAM Reactor that has been proposed as a power reactor by JAERI. The main requirements and the target values in the DREAM conceptual design were described, and compared to available experimental data. (author)

Design, manufacture and installation of measuring and control equipments for the advanced thermal prototype reactor 'Fugen'

International Nuclear Information System (INIS)

Hirota, Shigeo; Kawabata, Yoshinori

1979-01-01

The advanced thermal prototype reactor ''Fugen'' attained the criticality on March 20, 1978, and 100% output operation on November 13, 1978. On March 20, 1979, it passed the final inspection, and since then, it has continued the smooth operation up to now. The measuring and control equipments are provided for grasping the operational conditions of the plant and operating it safely and efficiently. At the time of designing, manufacturing and installing the measuring and control equipments for Fugen, it was required to clarify the requirements of the plant design, to secure the sufficient functions, and to improve the operational process, maintainability and the reliability and accuracy of the equipments. Many design guidelines and criteria were decided in order to coordinate the conditions among five manufacturers and give the unified state as one plant. The outline of the instrumentations for neutrons, radiation monitoring and process data, the control systems for reactivity, reactor output, pressure and water supply, the safety protection system, and the process computer are described. Finally, the installations and tests of the measuring and control equipments are explained. The aseismatic capability of the equipments was confirmed. (Kako, I.)

Experimental Facilities and Plan for a Prototype SFR

International Nuclear Information System (INIS)

Chang, Jinwook

2013-01-01

Contents of the presentation: 1. STELLA; 2. Under Sodium Viewing; 3. Sodium-CO 2 Interaction Test; Overview of the Sodium Integral Effect Test Loop for Safety Simulation and Assessment (STELLA) Program, the scope of the experiment and the overall characteristics of STELLA-1: Phase 1: STELLA-1, Individual component test; • Performance evaluation of key sodium components; • Heat exchanger design codes V&V. Phase 2: STELLA-2, Integral effect test; • Verification of dynamic plant response after reactor shutdown; • Construction of test DB to support specific design approval for the prototype SFR

Advances in U.S. reactor physics standards

International Nuclear Information System (INIS)

Cokinos, Dimitrios

2008-01-01

The standards for Reactor Design, widely used in the nuclear industry, provide guidance and criteria for performing and validating a wide range of nuclear reactor calculations and measurements. Advances, over the past decades in reactor technology, nuclear data and infrastructure in the data handling field, led to major improvements in the development and application of reactor physics standards. A wide variety of reactor physics methods and techniques are being used by reactor physicists for the design and analysis of modern reactors. ANSI (American National Standards Institute) reactor physics standards, covering such areas as nuclear data, reactor design, startup testing, decay heat and fast neutron fluence in the pressure vessel, are summarized and discussed. These standards are regularly undergoing review to respond to an evolving nuclear technology and are being successfully used in the U.S and abroad contributing to improvements in reactor design, safe operation and quality assurance. An overview of the overall program of reactor physics standards is presented. New standards currently under development are also discussed. (authors)

Reducing the cost of S-65C grade beryllium for ITER first wall applications

International Nuclear Information System (INIS)

Kaczynski, D.; Sato, K.; Savchuk, V.V.; Shestakov, V.P.

2004-01-01

Beryllium is the current material of choice for plasma-facing components in ITER. The present design is for 10 mm thick beryllium tiles bonded to an actively cooled copper substrate. Brush Wellman grade S65C beryllium is preferred grade off beryllium for these tiles. S65C has the best resistance to low-cycle thermal fatigue than any other beryllium grad in the world. S65C grade beryllium has been successfully deployed in fusion reactors for more than two decades, most recently in the JET reactor. This paper will detail a supply chain to produce the most cost-effective S65C plasma facing components for ITER. This paper will also propose some future work too demonstrate the best technology for bonding beryllium to copper. (author)

Start-up analysis of INET-5 MW district heating prototype reactor

International Nuclear Information System (INIS)

Li Tianshu

1991-09-01

The main features and thermohydraulic design parameters of the INET-5 MW reactor (INET: Institute of Nuclear Technology of Tsinghua University, Beijing) are presented. The start-up process and the effect of thermohydraulic instability on start-up process have been analyzed. The main obstacle of start-up process of INET-5 MW reactor is to pass the instability region from 1 atm to normal operation condition. For avoiding instability, the start-up process should be divided into two steps. The results of three different start-up proposals calculated by DACOL code are given and compared. The possibility of instabilities for each proposal has been checked. The checked results show that there is no instability during start-up of the three proposals. So, it is supposed that the INET-5 MW reactor can safely and stably reach the operation conditions. Finally, some conclusions about the effect of instability on start-up in boiling mode of INET-5MW reactor are given

A Computuerized Operator Support System Prototype

Energy Technology Data Exchange (ETDEWEB)

Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lew, Roger [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ulrich, Tom [Idaho National Lab. (INL), Idaho Falls, ID (United States); Villim, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-11-01

on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment.

A Computuerized Operator Support System Prototype

Energy Technology Data Exchange (ETDEWEB)

Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lew, Roger [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ulrich, Tom [Idaho National Lab. (INL), Idaho Falls, ID (United States); Villim, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-08-01

on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment.

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Experiment data report for semiscale MOD-1 tests S-03-A, S-03-B, S-03-C, and S-03-D (reflood heat transfer tests)

International Nuclear Information System (INIS)

1976-05-01

Recorded test data are presented for Tests S-03-A, S-03-B, S-03-C, and S-03-D of the Semiscale Mod-1 reflood heat transfer series (Test Series 3). The tests conducted in this series are separate effects core reflood tests performed to determine the reflood heat transfer characteristics of the 5.5 foot Mod-1 rod bundle. Tests S-03-A through S-03-D were forced-feed reflood tests in which the reflood rate was held constant during each test. The tests were conducted to investigate the effects on system response resulting from variations in operating conditions of pressure, temperature, core power, reflood coolant, subcooling, and peak heater rod thermocouple temperature at reflood initiation. Test S-03-A was conducted from an initial system temperature of about 230 0 F at a pressure of 20 psia. Tests S-03-B through S-03-D were conducted from an initial system temperature of about 290 0 F at a pressure of 60 psia. In all four tests, reflood coolant was injected directly into the core barrel by means of a specially designed core inlet manifold. The electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core during reflood. All four tests were conducted with a flat radial power profile. During reflood, core power was reduced from the initial level according to the American Nuclear Society (ANS) decay heat curve plus 20 percent for pressurized water reactor (PWR) core decay heat. The cold leg broken loop piping was open to the pressure suppression system (PSS). A separate steam supply system connected to the PSS was controlled to maintain constant pressure during the tests

Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

2016-05-15

KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

CO2 Reduction Assembly Prototype Using Microlith-Based Sabatier Reactor for Ground Demonstration

Science.gov (United States)

Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

2014-01-01

The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA's cabin Atmosphere Revitalization System (ARS) and In-Situ Resource Utilization (ISRU) architectures for both low-earth orbit and long-term manned space missions. Carbon dioxide can be reacted with H2, obtained from the electrolysis of water, via Sabatier reaction to produce methane and H2O. Methane can be stored and utilized as propellant while H2O can be either stored or electrolyzed to produce oxygen and regain the hydrogen atoms. Depending on the application, O2 can be used to replenish the atmosphere in human-crewed missions or as an oxidant for robotic and return missions. Precision Combustion, Inc. (PCI), with support from NASA, has previously developed an efficient and compact Sabatier reactor based on its Microlith® catalytic technology and demonstrated the capability to achieve high CO2 conversion and CH4 selectivity (i.e., =90% of the thermodynamic equilibrium values) at high space velocities and low operating temperatures. This was made possible through the use of high-heat-transfer and high-surface-area Microlith catalytic substrates. Using this Sabatier reactor, PCI designed, developed, and demonstrated a stand-alone CO2 Reduction Assembly (CRA) test system for ground demonstration and performance validation. The Sabatier reactor was integrated with the necessary balance-of-plant components and controls system, allowing an automated, single "push-button" start-up and shutdown. Additionally, the versatility of the test system prototype was demonstrated by operating it under H2-rich (H2/CO2 of >4), stoichiometric (ratio of 4), and CO2-rich conditions (ratio of <4) without affecting its performance and meeting the equilibrium-predicted water recovery rates. In this paper, the development of the CRA test system for ground demonstration will be discussed. Additionally, the performance results from testing the system at

Z-1 Prototype Space Suit Testing Summary

Science.gov (United States)

Ross, Amy

2013-01-01

The Advanced Space Suit team of the NASA-Johnson Space Center performed a series of test with the Z-1 prototype space suit in 2012. This paper discusses, at a summary level, the tests performed and results from those tests. The purpose of the tests were two-fold: 1) characterize the suit performance so that the data could be used in the downselection of components for the Z-2 Space Suit and 2) develop interfaces with the suitport and exploration vehicles through pressurized suit evaluations. Tests performed included isolated and functional range of motion data capture, Z-1 waist and hip testing, joint torque testing, CO2 washout testing, fit checks and subject familiarizations, an exploration vehicle aft deck and suitport controls interface evaluation, delta pressure suitport tests including pressurized suit don and doff, and gross mobility and suitport ingress and egress demonstrations in reduced gravity. Lessons learned specific to the Z-1 prototype and to suit testing techniques will be presented.

Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

Energy Technology Data Exchange (ETDEWEB)

Wood, RT

2004-09-27

This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

International Nuclear Information System (INIS)

Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

2005-01-01

G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

FFTF scale-model characterization of flow-induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36 percent to 111 percent of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable

FFTF scale-model characterization of flow induced vibrational response of reactor internals

Energy Technology Data Exchange (ETDEWEB)

Ryan, J A; Julyk, L J [Hanford Engineering Development Laboratory, Richland, WA (United States)

1977-12-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

FFTF scale-model characterization of flow induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

Core disruptive accident analysis in prototype fast breeder reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

2002-01-01

Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

LOCA simulation in the NRU reactor: materials test-1

International Nuclear Information System (INIS)

Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

1981-10-01

A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607 0 F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions

Space reactor fuel element testing in upgraded TREAT

International Nuclear Information System (INIS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

Space reactor fuel element testing in upgraded TREAT

Science.gov (United States)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

Science.gov (United States)

Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

Studies of cosmic ray events in ATLAS sTGC muon chamber prototypes

CERN Document Server

AUTHOR|(CDS)2097847; Warburton, Andreas

Four years after its first long shutdown in 2015, the Large Hadron Collider (LHC) will be shut down once more for a luminosity upgrade. During that time, the ATLAS detector on the LHC ring will also follow an upgrade program, one upgrade being the replacement of the Small Muon Wheels for a New Small Wheel containing small-strip Thin Gap Chambers (sTGCs). The sTGCs built in Canada will be tested at McGill University before their installation in ATLAS. A testing facility has been constructed and a 40 × 60 cm^2 sTGC prototype has been used to deliver preliminary measurements from cosmic rays. This thesis will present the development of a robust tracking algorithm which can handle extra clusters and multiple tracks in an sTGC detector. This algorithm also categorizes events based on their number of clusters and tracks. By modifying the trigger time window of the sTGC prototype, the evolution of the distribution of events over this categorization is shown.

Advanced reactor development

International Nuclear Information System (INIS)

Till, C.E.

1989-01-01

Consideration is given to what the aims of advanced reactor development have to be, if a new generation of nuclear power is really to play an important role in man's energy generation activities in a fragile environment. The background given briefly covers present atmospheric evidence, the current situation in nuclear power, how reactors work and what can go wrong with them, and the present magnitudes of world energy generation. The central part of the paper describes what is currently being done in advanced reactor development and what can be expected from various systems and various elements of it. A vigorous case is made that three elements must be present in any advanced reactor development: (1) breeding; (2) passive safety; and (3) shorter-live nuclear waste. All three are possible. In the right advanced reactor systems the ways of achieving them are known. But R and D is necessary. That is the central argument made in the paper. Not advanced reactor prototype construction at this point, but R and D itself. (author)

Solution Prototype

DEFF Research Database (Denmark)

Efeoglu, Arkin; Møller, Charles; Serie, Michel

2013-01-01

This paper outlines an artifact building and evaluation proposal. Design Science Research (DSR) studies usually consider encapsulated artifact that have relationships with other artifacts. The solution prototype as a composed artifact demands for a more comprehensive consideration in its systematic...... environment. The solution prototype that is composed from blending product and service prototype has particular impacts on the dualism of DSR’s “Build” and “Evaluate”. Since the mix between product and service prototyping can be varied, there is a demand for a more agile and iterative framework. Van de Ven......’s research framework seems to fit this purpose. Van de Ven allows for an iterative research approach to problem solving with flexible starting point. The research activity is the result between the iteration of two dimensions. This framework focuses on the natural evaluation, particularly on ex...

The prototype of radioactive ion source

CERN Document Server

Aleksandrov, A V; Kot, N K; Andrighetto, A; Stroe, L

2001-01-01

The design and experimental results of the RIB source prototype are presented.A source will have the container of sup 2 sup 3 sup 5 U compounds heated up to 2200-2500 degree C. Vapors of uranium fission obtained when the ion source is irradiated by the high-energy neutron flux, are then ionized and extracted from the source. In the experiments with the prototype loaded by sup 1 sup 2 C the source working temperature 2700 degree C was reached, the carbon ion current 10 nA was obtained. The total operation time of more than 100 hours with no performance degradation was demonstrated.

Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

International Nuclear Information System (INIS)

Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

2009-01-01

Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

Study on the Υ(1S)→B_cM Weak Decays

International Nuclear Information System (INIS)

Huang, Jinshu; Chang, Qin; Wang, Na; Chen, Lili; Sun, Junfeng; Yang, Yueling

2015-01-01

Motivated by the prospects of the potential Υ(1S) particle at high-luminosity heavy-flavor experiments, we studied the Υ(1S)→B_cM weak decays, where M = π, ρ, K"("∗"). The nonfactorizable contributions to hadronic matrix elements are taken into consideration with the QCDF approach. It is found that the CKM-favored Υ(1S)→B_cρ decay has branching ratio of O(10"-"1"0), which might be measured promisingly by the future experiments.

RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

International Nuclear Information System (INIS)

Jumel, St.

2005-01-01

The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

Development of Korea advanced liquid metal reactor

International Nuclear Information System (INIS)

Park, C.K.

1998-01-01

Future nuclear power plants should not only have the features of improved safety and economic competitiveness but also provide a means to resolve spent fuel storage problems by minimizing volume of high level wastes. It is widely believed that liquid metal reactors (LMRs) have the highest potential of meeting these requirements. In this context, the LMR development program was launched as a national long-term R and D program in 1992, with a target to introduce a commercial LMR around 2030. Korea Advanced Liquid Metal Reactor (KALIMER), a 150 MWe pool-type sodium cooled prototype reactor, is currently under the conceptual design study with the target schedule to complete its construction by the mid-2010s. This paper summarizes the KALIMER development program and major technical features of the reactor system. (author)

Design evaluation of the semi-prototype for the ITER blanket first wall qualification

International Nuclear Information System (INIS)

Lee, Dong Won; Bae, Young Dug; Kim, Suk Kwon; Kim, Sun Ho; Hong, Bong Guen; Bang, In Cheol

2010-01-01

For the second qualification of the First Wall (FW) procurement of the International Thermonuclear Experimental Reactor (ITER), a semi-prototype of the FW has been designed with increased local surface heat flux up to 5 MW/m 2 . With the given conditions, the new semi-prototype design was simulated with the commercial CFD code, the ANSYS-11. The results show that the semi-prototype temperature exceeds the melting temperature of Be, and the current design is required to be modified. In order to enhance cooling, a hypervapotron was added in the design and an analysis with the same code was performed. The results show that the temperature with the hypervapotron reduced by around 100 o C but it was still higher than the melting temperature of Be. The hypervapotron mock-up was fabricated and tested with a variance of inlet coolant flow rates and heat fluxes of up to 1.75 MW/m 2 using the second Korea Heat Load Test (KoHLT-2) facility, in which heat was loaded by a graphite heater through radiation heating. Wall and coolant temperatures were measured and compared with the simulation results. So far, there is a large difference between the experiments and the simulation, and a next experiment is being prepared.

Modernization and Refurbishment of the RECH-1 Reactor

Energy Technology Data Exchange (ETDEWEB)

Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

2014-08-15

The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

Identification of Chemical Reactor Plant’s Mathematical Model

OpenAIRE

Pyakullya, Boris Ivanovich; Kladiev, Sergey Nikolaevich

2015-01-01

This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

C.I.S. natural gas-1

International Nuclear Information System (INIS)

Carson, M.; Stram, B.

1993-01-01

This paper reports that in the countries that make up the Commonwealth of Independent States (C.I.S.), with their vast resources and a considerable existing production base, prospects are good for further growth of the region's exportable gas surplus. Investment fundamentals are stronger for gas than for any other energy resources in the area. But the pipeline infrastructure to move large amounts of gas will need extensive refurbishment to ensure export reliability and growth. Given the potential in terms of production and markets, significant amounts of outside investment in oil, natural gas, and NGL infrastructure will likely increase dramatically in these countries in the near future. These are some of the major conclusions of Enron Corp.'s recent investigations in the C.I.S. and other former Soviet republics

A CAREM fuel assembly prototype construction in order to verify its mechanical design using hydrodynamic testing

International Nuclear Information System (INIS)

Aparicio, Gaspar; Di Marco, Agustin; Falcone, Jose M.; Giorgis, Miguel A.; Mathot, Sergio R.; Migliori, Julio; Orlando, Oscar S.; Restelli, Miguel A.; Ruggirello, Gabriel; Sapia, Gustavo C.; Zinzallari, Fausto; Bianchi, Daniel R.; Volpi, Ricardo M.

2000-01-01

The scope of this paper is to describe the activities of several Groups from three Atomic Centers (C. A. Bariloche, C. A. Ezeiza and C. A. Constituyentes), involved in the manufacturing of a CAREM fuel assembly prototype. The Design Group (UAIN-CAB) carried out the fuel assembly engineering. Cladding components were constructed by the Special Alloys Pilot Factory (UAMCN-CAE). Engineering Group (UACN-CAC) manufactured the parts to be processed, resorting to qualified suppliers. Elastic spacers were completely designed and constructed by this Group, and fuel rods, control rods, guide tubes and spacers were also welded here. Research Reactors Fuels Group (UACN-CAC) carried out the dimensional control of the elaborated parts, while Postirradiation Testing Group (UACN-CAC) performed the assembling of the fuel element. This paper also refers to the design and development of special equipment and devices, all of them required for the prototype construction. (author)

Radionuclide trap for liquid metal cooled reactors

International Nuclear Information System (INIS)

McGuire, J.C.; Brehm, W.F.

1978-10-01

At liquid metal cooled reactor operating temperatures, radioactive corrosion product transport and deposition in the primary system will be sufficiently high to limit access time for maintenance of system components. A radionuclide trap has been developed to aid in controlling radioactivity transport. This is a device which is located above the reactor core and which acts as a getter, physically immobilizing radioactive corrosion products, particularly 54 Mn. Nickel is the getter material used. It is most effective at temperatures above 450 0 C and effectiveness increases with increasing temperature. Prototype traps have been tested in sodium loops for 40,000 hours at reactor primary temperatures and sodium velocities. Several possible in-reactor trap sites were considered but a location within the top of each driver assembly was chosen as the most convenient and effective. In this position the trap is changed each time fuel is changed

Prototype thermochemical heat storage with open reactor system

NARCIS (Netherlands)

Zondag, H.A.; Kikkert, B.; Smeding, S.F.; Boer, de R.; Bakker, M.

2013-01-01

Thermochemical (TC) heat storage is an interesting technology for future seasonal storage of solar heat in the built environment. This technology enables high thermal energy storage densities and low energy storage losses. A small-scale laboratory prototype TC storage system has been realized at

[Present conceptions of the C.E.A. concerning] the development of fast neutron reactors in France

International Nuclear Information System (INIS)

Vendryes, G.; Gaussens, J.

1964-01-01

1 - The position of fast neutron reactors in the French nuclear energy program. In developing a program based on natural uranium, France will have an important stock of plutonium rich in higher isotopes. The existence of this plutonium and of the depleted uranium arising from the same reactors, has, as a logical consequence, the use of both in fast neutron reactors. Justified by this short term interest, the achievement of fast neutron reactors does, moreover, provide for a future necessity. 2 - Description of a fast neutron central power station of 1000 MWe. We indicate the characteristics of a future fast neutron central power station, plutonium fuelled, and sodium cooled. However uncertain these characteristics may be, they constitute a necessary guide in the orientation of our work. 3 - Studies carried out up to the present time. We give an outline of those studies, often very preliminary, which have given the characteristics cited above. The principal technical areas taken up are the following: - Neutronics (critical masses, breeding ratios, enrichments, flattening of the neutron flux, coefficients of reactivity, reactivity changes as a function of irradiation). - Dynamics, control, and safety. - Technology (design of the core and vessel, of the sodium system, and of the fuel handling mechanisms). These technical studies are complemented by economic considerations. The choice of the optimum characteristics is related to the existence of power production programs, and, in these programs, to the existence of plutonium producing thermal reactors. It is shown how, in this context, the existence of plutonium should be taken into account, and, in addition which mechanisms relate the economics of this plutonium to the choice of the most important parameters of the breeder reactors. 4 - Prototype reactor. The interest in an intermediate stage consisting of a reactor of a power level of about 80 MWe is justified. Its essential characteristics are briefly presented

Remote Reactor Monitoring

Energy Technology Data Exchange (ETDEWEB)

Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-10-21

The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Status of advanced nuclear reactor development in Korea

International Nuclear Information System (INIS)

Kim, H.R.; Kim, K.K.; Kim, Y.W.; Joo, H.K.

2014-01-01

The Korean nuclear industry is facing new challenges to solve the spent fuel storage problem and meet the needs to diversify the application areas of nuclear energy. In order to provide solutions to these challenges, the Korea Atomic Energy Research Institute (KAERI) has been developing advanced nuclear reactors including a Sodium-cooled Fast Reactor, Very High Temperature Gas cooled Reactor (VHTR), and System-integrated Modular Advanced Reactor (SMART) with substantially improved safety, economics, and environment-friendly features. A fast reactor system is one of the most promising options for a reduction of radioactive wastes. The long-term plan for Advanced SFR development in conjunction with the pyro-process was authorized by the Korean Atomic Energy Commission in 2008. The development milestone includes specific design approval of a prototype SFR by 2020, and the construction of a prototype SFR by 2028. KAERI has been carrying out the preliminary design of a 150MWe SFR prototype plant system since 2012. The development of advanced SFR technologies and the basic key technologies necessary for the prototype SFR are also being carried out. By virtue of high-temperature heat, a VHTR has diverse applications including hydrogen production. KAERI launched a nuclear hydrogen project using a VHTR in 2006, which focused on four basic technologies: the development of design tools, very high-temperature experimental technology, TRISO fuel fabrication, and Sulfur-iodine thermo-chemical hydrogen production technology. The technology development project will be continued until 2017. A conceptual reactor design study was started in 2012 as collaboration between industry and government to enhance the early-launching of the nuclear hydrogen development and demonstration (NHDD) project. The goal of the NHDD project is to design and build a nuclear hydrogen demonstration system by 2030. KAERI has developed SMART which is a small-sized advanced integral reactor with a rated

Characterization of 2D-C/C composite for application of very high temperature reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sumita, Junya; Kunimoto, Eiji; Sawa, Kazuhiro; Makita, Taiyo; Takagi, Takashi; Kim, W.J.; Jung, C.H.; Park, J.Y.

2010-01-01

For in-core components of VHTR (Very High Temperature Reactor), carbon fiber reinforced carbon matrix composite (C/C composite) is one of the major candidate materials. In this study, fracture behaviors of two dimensional (2D-) C/C composites were examined by SENB specimens with four-point bending test. The surface of specimens was observed by a CCD camera during the bending test, and observed by a stereomicroscope before and after the bending test. The following results were obtained through mode-I fracture test. (1) Three types of the composites were evaluated by tentatively using the stress intensity factor equation for metallic materials. The equivalent stress intensity factor of 2D-C/C composite is in the range of 5.9 - 10.0MPa m 1/2 . It was expected that the fracture mechanism for the composite materials could be assessed by this test method. (2) The crack opening displacement-load behavior of C/C composite might depend not only on the propagation of crack but also on delaminating between layers. (author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected. (author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected

Modifications done in the IPR-R1 reactor and their auxiliary systems

International Nuclear Information System (INIS)

Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

1986-01-01

The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

Identification of Chemical Reactor Plant’s Mathematical Model

Directory of Open Access Journals (Sweden)

Pyakillya Boris

2015-01-01

Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

Implementation of new core cooling monitoring system for light water reactors - BCCM (Becker Core Cooling Monitor)

International Nuclear Information System (INIS)

Coville, Patrick; Eliasson, Bengt; Stromqvist, Erik; Ward, Olav; Fox, Georges; Ashjian, D. T.

1998-01-01

Core cooling monitors are key instruments to protect reactors from large accidents due to loss of coolant. Sensors presented here are based on resistance thermometry. Temperature dependent resistance is powered by relatively high and constant current. Value of this resistance depends on thermal exchange with coolant and when water is no more surrounding the sensors a large increase of temperature is immediately generated. The same instrument can be operated with low current and will measure the local temperature up to 1260 o C in case of loss of coolant accident. Sensors are manufactured with very few components and materials already qualified for long term exposure to boiling or pressurized water reactors environment. Prototypes have been evaluated in a test loop up to 160 bars and in the Barsebaeck-1 reactor. Industrial sensors are now in operation in reactor Oskarshamn 2. (author)

Hydraulic experiments on the failed fuel location module of prototype fast breeder reactor

International Nuclear Information System (INIS)

Rajesh, K.; Kumar, S.; Padmakumar, G.; Prakash, V.; Vijayashree, R.; Rajan Babu, V.; Govinda Rajan, S.; Vaidyanathan, G.; Prabhaker, R.

2003-01-01

The design of Prototype Fast Breeder Reactor (PFBR) is based on sound design concepts with emphasis on intrinsic safety. The uncertainties involved in the design of various components, which are difficult to assess theoretically, are experimentally verified before design is validated. In PFBR core, the coolant (liquid sodium) enters the bottom of the fuel subassembly, passes over the fuel pins picking up the fission heat and issues in to a hot pool. If there is any breach in the fuel pins, the fission products come in direct contact with the coolant. This is undesirable and it is necessary to locate the subassembly with the failed fuel pin and to isolate it. A component called Failed Fuel Location Module (FFLM) is employed for locating the failed SA by monitoring the coolant samples coming out of each Subassembly. The coolant sample from each Subassembly is drawn by FFLM using an EM pump through sampling tube and selector valve and is monitored for the presence of delayed neutrons which is an indication of failure of the Subassembly. The pressure drop across the selector valve determines the rating of the EM Pump. The dilution of the coolant sample across the selector valve determines the effectiveness of monitoring for contamination. It is not possible to predict pressure drop across the selector valve and dilution of the coolant sample theoretically. These two parameters are determined using a hydraulic experiment on the FFLM. The experiment was carried out in conditions that simulate the reactor conditions following appropriate similarity laws. The paper discusses the details of the model, techniques of experiments and the results from the studies

Technical report on implementation of reactor internal 3D modeling and visual database system

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

Power from plutonium: fast reactor fuel

International Nuclear Information System (INIS)

Bishop, J.F.W.

1981-01-01

Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

U.S., European ALMA Partners Award Prototype Antenna Contracts

Science.gov (United States)

2000-03-01

The U.S. and European partners in the Atacama Large Millimeter Array (ALMA) project have awarded contracts to U.S. and Italian firms, respectively, for two prototype antennas. ALMA is a planned telescope array, expected to consist of 64 millimeter-wave antennas with 12-meter diameter dishes. The array will be built at a high-altitude, extremely dry mountain site in Chile's Atacama desert, and is scheduled to be completed sometime in this decade. On February 22, 2000, Associated Universities Inc. (AUI) signed an approximately $6.2 million contract with Vertex Antenna Systems, of Santa Clara, Calif., for construction of one prototype ALMA antenna. AUI operates the U.S. National Radio Astronomy Observatory (NRAO) for the National Science Foundation under a cooperative agreement. The European partners contracted with the consortium of European Industrial Engineering and Costamasnaga, of Mestre, Italy, on February 21, 2000, for the production of another prototype. (Mestre is located on the inland side of Venice.) The two antennas must meet identical specifications, but will inherently be of different designs. This will ensure that the best possible technologies are incorporated into the final production antennas. Only one of the designs will be selected for final production. Several technical challenges must be met for the antennas to perform to ALMA specifications. Each antenna must have extremely high surface accuracy (25 micrometers, or one-third the diameter of a human hair, over the entire 12-meter diameter). This means that, when completed, the surface accuracy of the ALMA dishes will be 20 times greater than that of the Very Large Array (VLA) antennas, and about 50 times greater than dish antennas for communications or radar. The ALMA antennas must also have extremely high pointing accuracy (0.6 arcseconds). An additional challenge is that the antennas, when installed at the ALMA site in Chile, will be exposed to the ravages of weather at 16,500 feet (5000 meters

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

University Reactor Matching Grants Program

International Nuclear Information System (INIS)

John Valentine; Farzad Rahnema; Said Abdel-Khalik

2003-01-01

During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given

Operating history of U.S. nuclear power reactors

International Nuclear Information System (INIS)

1974-01-01

The operating history of U. S. nuclear power plants through December 31, 1974 has been collected. Included are those nuclear reactor facilities which produce electricity, even if in token amounts, or which are part of a development program concerned with the generation of electricity through the use of a nuclear reactor as a heat source. The information is based on data furnished by facility operators. The charts are plotted in terms of cumulative thermal energy as a function of time. Since only those shutdowns of five days or more are shown, the charts do not give a detailed history of plant operation. They do, however, give an overview of the operating history of a variety of developmental and experimental nuclear power reactors. The data show the yearly gross generation of electricity for each U. S. nuclear plant and, for civilian power plants, information on reactor availability and plant capacity factor. (U.S.)

SiC epitaxial layer growth in a novel multi-wafer VPE reactor

Energy Technology Data Exchange (ETDEWEB)

Burk, A.A. Jr.; O`Loughlin, M.J. [Northrop Grumman Advanced Technology Lab., Baltimore, MD (United States); Mani, S.S. [Northrop Grumman Science and Technology Center, Pittsburgh, PA (United States)

1998-06-01

Preliminary results are presented for SiC epitaxial layer growth employing a unique planetary SiC-VPE reactor. The high-throughput, multi-wafer (7 x 2-inch) reactor, was designed for atmospheric and reduced pressure operation at temperatures up to and exceeding 1600 C. Specular epitaxial layers have been grown in the reactor at growth rates from 3-5 {mu}m/hr. The thickest layer grown to data was 42 {mu}m. The layers exhibit minimum unintentional n-type doping of {proportional_to}1 x 10{sup 15} cm{sup -3}, room temperature mobilities of {proportional_to}1000 cm{sup 2}/Vs, and intentional n-type doping from {proportional_to}5 x 10{sup 15} cm{sup -3} to >1 x 10{sup 19} cm{sup -3}. Intrawafer thickness and doping uniformities of 4% and 7% (standard deviation/mean) have been obtained, respectively, on 35 mm diameter substrates. Recently, 3% thickness uniformity has been demonstrated on a 50 mm substrate. Within a run, wafer-to-wafer thickness deviation is {proportional_to}4-14%. Doping variation is currently larger, ranging as much as a factor of two from the highest to the lowest doped wafer. Continuing efforts to improve the susceptor temperature uniformity and reduce unintentional hydrocarbon generation to improve layer uniformity and reproducibility, are presented. (orig.) 18 refs.

Research about reactor operator's personability characteristics and performance

Energy Technology Data Exchange (ETDEWEB)

Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

2003-03-01

To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

Review of fast reactor activities in India

International Nuclear Information System (INIS)

Paranjpe, S.R.

1982-01-01

A review of fast reactor activities in India is introduced. One stage of construction of the Fast Breeder Test Reactor (FBTR) and design studies for 500MWe Prototype Fast Breeder Reactor (PFBR) are briefly summarized. The emphasis is on fast reactor physics, materials studies, radiochemistry, and the safety and fuel reprocessing programme

Outline of design, manufacturing and installation experience of pressure vessel structure for the prototype heavy water moderated boiling light water cooled reactor 'FUGEN'

International Nuclear Information System (INIS)

Shibato, Eizo; Oguchi, Isao; Kishi, Toshikazu; Kitagawa, Yuji

1977-01-01

After component installation completed in June 1977 and various functional tests to be conducted later, the prototype heavy water moderated, boiling light water cooled reactor ''FUGEN'' is scheduled to reach first criticality in March 1978. Since the pressure vessel of ''FUGEN'' is completely different from that of a light water reactor in structure and materials, through research and development work was carried out prior to fabrication and construction. Based on these studies, installation of the actual pressure vessel was completed. Functional tests are now under way. This article describes examples in which our research and development results are reflected on design, manufacture, and installation of the pressure vessel. Also it introduces noteworthy achievements relevant to production techniques in manufacture and installation. (auth.)

Calculation of the source term for a S1B-sequence at a VVER-1000 type reactor. Part 1

International Nuclear Information System (INIS)

Sdouz, G.

1990-10-01

The behaviour of the source term in a VVER-1000 type reactor is calculated using the 'Source Term Code Package' (STCP). The input data are based on the russian plant Zaporozhye-5. The selected accident sequence is a small break LOCA in the hot leg followed by loss offsite and onsite electric power (S 1 B-sequence). According to the course of the calculation the results are presented and analyzed for each program. Except for the noble gases all release fractions are lower than 10 -4 . 18 refs., 10 tabs. (Author)

A review of the UK fast reactor programme. March 1977

International Nuclear Information System (INIS)

Smith, R.D.

1977-01-01

This paper reports on the Fast Reactor Programme of United Kingdom. These are the main lines: Dounreay Fast Reactor; Prototype Fast Reactor; Commercial Fast Reactor; engineering development; materials development; chemical engineering/sodium technology; fast reactor fuel; fuel cycle; safety; reactor performance study

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han

2013-01-01

The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.

11 CFR 101.1 - Candidate designations (2 U.S.C. 432(e)(1)).

Science.gov (United States)

2010-01-01

... 11 Federal Elections 1 2010-01-01 2010-01-01 false Candidate designations (2 U.S.C. 432(e)(1)). 101.1 Section 101.1 Federal Elections FEDERAL ELECTION COMMISSION GENERAL CANDIDATE STATUS AND... and address, party affiliation, and office sought, the District and State in which Federal office is...

Safe decommissioning of the Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Garlea, C.; Garlea, I.; Kelerman, C.; Rodna, A.

2002-01-01

The VVR-S Romania research reactor was operated between 1957-1997, at 2 MW nominal power, for research and radioisotopical production. The detailed decommissioning plan was developed between 1995-1998, in the frame of the International Atomic Energy Agency Technical assistance project ROM/9/017. The proposed strategy agreed by the counterpart as well as international experts was stage 1. In 1997, an independent analysis performed by European Commission experts, in the frame of PHARE project PH04.1/1994 was dedicated to the 'Study of Soviet Design Research Reactors', had consolidated the development of the project emphasizing technical options of safe management for radioactive wastes and VVR-S spent fuel. The paper presents the main technical aspects as well as those of social impact, which lead to the establishment of strategy for safe management of decommissioning. Technical analysis of the VVR-S reactor and associated radwaste facilities (Radioactive Waste Treatment Plant - Magurele and National Repository Baita-Bihor) proved the possibility of the classical method utilization for dismantling of the facility and treatment-conditioning-disposal of the arrised wastes in safe conditions. The decommissioning plan at stage 2 has been developed based on radiological safety assessment, evaluation of radwaste inventory (removed as well as preserved on site), cost analysis and environmental impact. Technical data were provided by the R and D programme including neutron calculations and experiments, radiological characterizing (for facility and its influence area), seismic analysis and environmental balance during the operation and after shut down of the reactor. A special chapter is dedicated to regulatory issues concerning the development of decommissioning under nuclear safety. Based on the Fundamental Norms of Radiological Safety, the Regulatory Body defined the clearance levels and safety criteria for the process. The development of National Norms for the

Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-08-15

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Sang June Ahn

2016-08-01

Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

Status of fast reactor control rod development in the United Kingdom

International Nuclear Information System (INIS)

Kelly, B.T.

1984-01-01

The two large fast reactors constructed in the United Kingdom, that is the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR) differed substantially in their control systems. DFR was controlled by variation of the neutron leakage from the core while PFR uses conventional control rods containing neutron absorbing materials. This paper describes the development of the PFR control systems, the progressive design of the control systems for the prototype Civil Fast Reactor (CFR) and the supporting research and development programmes. (author)

Estimations on uranium silicide fuel prototypes for their irradiation and postirradiation

International Nuclear Information System (INIS)

Sbaffoni, Maria M.

2000-01-01

The 'Silicide' project includes the qualification of this type of research reactor fuel to be used i.e. in the Argentine RA-3 and to confirm CNEA's role as an international supplier. The present paper shows complementary basic information for P-04 prototype post-irradiation, which is already under way, and some parameter values related to the new P-06 prototype to be taken into account for planning its irradiation and post-irradiation. The reliability of these values has been evaluated through comparison with experimental results. The reported results contribute, also, to a parallel study on the nuclear data libraries used in calculations for this type of reactor. (author)

Observation of $S=+1$ Narrow Resonances in the System $pK^0_s$ from $p+\\rm {C_3H_8}$ Collision at 10 GeV/$c$

CERN Document Server

Aslanyan, P Zh; Rikhvitskaya, G G

2004-01-01

Experimental data from a 2 m propane bubble chamber have been analyzed to search for an exotic baryon state, the $\\Theta^+$ baryon, in the $pK^0_s$ decay mode for the reaction $p+{\\rm C_3H_8}$ at 10 GeV/$c$. The $pK^0_s$ invariant mass spectrum shows resonant structures with $M_{p K_s^0}=1540\\pm 8$, $1613\\pm10$, $1821\\pm11$ MeV/$c^2$ and $\\Gamma_{p K_s^0}= 9.2\\pm1.8$, $16.1\\pm4.1$, $28.0\\pm9.4$ MeV/$c^2$. The statistical significance of these peaks has been estimated as $5.5$, $4.8$ and $5.0$ s.d., respectively. There are also small peaks in mass regions of 1487 (3.0 s.d.), 1690 (3.6 s.d.) and 1980 (3.0 s.d.) MeV/$c^2$.

Heat transfer in a magnet C

International Nuclear Information System (INIS)

Sircilli Neto, F.; Passaro, A.; Borges, E.M.

1991-01-01

The cooling systems of nuclear reactors for spacial applications include direct current electromagnetic pumps, which are used to circulate the coolant fluid thru the reactor core. In this work, the transfer of the heat generated by the electrical current in a magnet C excitation coils, which is used in a prototype pump, was evaluated. Considering the processes of heat transfer by conduction, natural convection and radiation, the results of simulation with the codes HEATING5 and AUTHEATS indicate the utilization of the 180 sup(0)C thermal class conductor for a working Joule power of 4 10 sup(4) W/m sup(3) in each magnet coil. (author)

Comparative prediction of irradiation test of CNFT and Cise prototypes of CIRENE fuel pins, a prediction by transuranus M1V1J12 code

International Nuclear Information System (INIS)

Suwardi

2014-01-01

A prototype of fuel pin design for HWR by CIRENE has been realized by Center for Nuclear Fuel Technology CNFT-BATAN. The prototype will be irradiated in PRTF Power Ramp Test (PRTF). The facility has been installed inside RSG-GA Siwabessy at Serpong. The present paper reports the preparation of experimentation and prediction of irradiation test. One previous PCI test report is found in, written by Lysell G and Valli G in 1973. The CNFT fuel irradiation test parameter is adapted to both PRTF and power loop design for RSG-GAS reactor in Serpong mainly the maxima of: rod length, neutrons flux, total power of rod, and power ramp rate. The CNFT CIRENE prototype design has been reported by Futichah et al 2007 and 2010. The AEC-India HWR fuel pin is of 19/22 fuel bundle design has also been evaluated as comparison. The first PCI test prediction has experiment comparison for Cise pin. The second prediction will be used for optimizing the design of ramp test for CNFT CIRENE fuel pin prototype. (author)

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Measurement of sigma chi c2 B(chi c2-->J/psi gamma)/sigma chi c1 B(chi c1 -->J/psi gamma) in pp collisions at square root s=1.96 TeV.

Science.gov (United States)

Abulencia, A; Adelman, J; Affolder, T; Akimoto, T; Albrow, M G; Ambrose, D; Amerio, S; Amidei, D; Anastassov, A; Anikeev, K; Annovi, A; Antos, J; Aoki, M; Apollinari, G; Arguin, J-F; Arisawa, T; Artikov, A; Ashmanskas, W; Attal, A; Azfar, F; Azzi-Bacchetta, P; Azzurri, P; Bacchetta, N; Badgett, W; Barbaro-Galtieri, A; Barnes, V E; Barnett, B A; Baroiant, S; Bartsch, V; Bauer, G; Bedeschi, F; Behari, S; Belforte, S; Bellettini, G; Bellinger, J; Belloni, A; Benjamin, D; Beretvas, A; Beringer, J; Berry, T; Bhatti, A; Binkley, M; Bisello, D; Blair, R E; Blocker, C; Blumenfeld, B; Bocci, A; Bodek, A; Boisvert, V; Bolla, G; Bolshov, A; Bortoletto, D; Boudreau, J; Boveia, A; Brau, B; Brigliadori, L; Bromberg, C; Brubaker, E; Budagov, J; Budd, H S; Budd, S; Budroni, S; Burkett, K; Busetto, G; Bussey, P; Byrum, K L; Cabrera, S; Campanelli, M; Campbell, M; Canelli, F; Canepa, A; Carillo, S; Carlsmith, D; Carosi, R; Carron, S; Casarsa, M; Castro, A; Catastini, P; Cauz, D; Cavalli-Sforza, M; Cerri, A; Cerrito, L; Chang, S H; Chen, Y C; Chertok, M; Chiarelli, G; Chlachidze, G; Chlebana, F; Cho, I; Cho, K; Chokheli, D; Chou, J P; Choudalakis, G; Chuang, S H; Chung, K; Chung, W H; Chung, Y S; Ciljak, M; Ciobanu, C I; Ciocci, M A; Clark, A; Clark, D; Coca, M; Compostella, G; Convery, M E; Conway, J; Cooper, B; Copic, K; Cordelli, M; Cortiana, G; Crescioli, F; Cuenca Almenar, C; Cuevas, J; Culbertson, R; Cully, J C; Cyr, D; DaRonco, S; Datta, M; D'Auria, S; Davies, T; D'Onofrio, M; Dagenhart, D; de Barbaro, P; De Cecco, S; Deisher, A; De Lentdecker, G; Dell'Orso, M; Delli Paoli, F; Demortier, L; Deng, J; Deninno, M; De Pedis, D; Derwent, P F; Di Giovanni, G P; Dionisi, C; Di Ruzza, B; Dittmann, J R; DiTuro, P; Dörr, C; Donati, S; Donega, M; Dong, P; Donini, J; Dorigo, T; Dube, S; Efron, J; Erbacher, R; Errede, D; Errede, S; Eusebi, R; Fang, H C; Farrington, S; Fedorko, I; Fedorko, W T; Feild, R G; Feindt, M; Fernandez, J P; Field, R; Flanagan, G; Foland, A; Forrester, S; Foster, G W; Franklin, M; Freeman, J C; Furic, I; Gallinaro, M; Galyardt, J; Garcia, J E; Garberson, F; Garfinkel, A F; Gay, C; Gerberich, H; Gerdes, D; Giagu, S; Giannetti, P; Gibson, A; Gibson, K; Gimmell, J L; Ginsburg, C; Giokaris, N; Giordani, M; Giromini, P; Giunta, M; Giurgiu, G; Glagolev, V; Glenzinski, D; Gold, M; Goldschmidt, N; Goldstein, J; Golossanov, A; Gomez, G; Gomez-Ceballos, G; Goncharov, M; González, O; Gorelov, I; Goshaw, A T; Goulianos, K; Gresele, A; Griffiths, M; Grinstein, S; Grosso-Pilcher, C; Group, R C; Grundler, U; Guimaraes da Costa, J; Gunay-Unalan, Z; Haber, C; Hahn, K; Hahn, S R; Halkiadakis, E; Hamilton, A; Han, B-Y; Han, J Y; Handler, R; Happacher, F; Hara, K; Hare, M; Harper, S; Harr, R F; Harris, R M; Hartz, M; Hatakeyama, K; Hauser, J; Heijboer, A; Heinemann, B; Heinrich, J; Henderson, C; Herndon, M; Heuser, J; Hidas, D; Hill, C S; Hirschbuehl, D; Hocker, A; Holloway, A; Hou, S; Houlden, M; Hsu, S-C; Huffman, B T; Hughes, R E; Husemann, U; Huston, J; Incandela, J; Introzzi, G; Iori, M; Ishizawa, Y; Ivanov, A; Iyutin, B; James, E; Jang, D; Jayatilaka, B; Jeans, D; Jensen, H; Jeon, E J; Jindariani, S; Jones, M; Joo, K K; Jun, S Y; Jung, J E; Junk, T R; Kamon, T; Karchin, P E; Kato, Y; Kemp, Y; Kephart, R; Kerzel, U; Khotilovich, V; Kilminster, B; Kim, D H; Kim, H S; Kim, J E; Kim, M J; Kim, S B; Kim, S H; Kim, Y K; Kimura, N; Kirsch, L; Klimenko, S; Klute, M; Knuteson, B; Ko, B R; Kondo, K; Kong, D J; Konigsberg, J; Korytov, A; Kotwal, A V; Kovalev, A; Kraan, A C; Kraus, J; Kravchenko, I; Kreps, M; Kroll, J; Krumnack, N; Kruse, M; Krutelyov, V; Kubo, T; Kuhlmann, S E; Kuhr, T; Kusakabe, Y; Kwang, S; Laasanen, A T; Lai, S; Lami, S; Lammel, S; Lancaster, M; Lander, R L; Lannon, K; Lath, A; Latino, G; Lazzizzera, I; LeCompte, T; Lee, J; Lee, J; Lee, Y J; Lee, S W; Lefèvre, R; Leonardo, N; Leone, S; Levy, S; Lewis, J D; Lin, C; Lin, C S; Lindgren, M; Lipeles, E; Lister, A; Litvintsev, D O; Liu, T; Lockyer, N S; Loginov, A; Loreti, M; Loverre, P; Lu, R-S; Lucchesi, D; Lujan, P; Lukens, P; Lungu, G; Lyons, L; Lys, J; Lysak, R; Lytken, E; Mack, P; MacQueen, D; Madrak, R; Maeshima, K; Makhoul, K; Maki, T; Maksimovic, P; Malde, S; Manca, G; Margaroli, F; Marginean, R; Marino, C; Marino, C P; Martin, A; Martin, M; Martin, V; Martínez, M; Maruyama, T; Mastrandrea, P; Masubuchi, T; Matsunaga, H; Mattson, M E; Mazini, R; Mazzanti, P; McFarland, K S; McIntyre, P; McNulty, R; Mehta, A; Mehtala, P; Menzemer, S; Menzione, A; Merkel, P; Mesropian, C; Messina, A; Miao, T; Miladinovic, N; Miles, J; Miller, R; Mills, C; Milnik, M; Mitra, A; Mitselmakher, G; Miyamoto, A; Moed, S; Moggi, N; Mohr, B; Moore, R; Morello, M; Movilla Fernandez, P; Mülmenstädt, J; Mukherjee, A; Muller, Th; Mumford, R; Murat, P; Nachtman, J; Nagano, A; Naganoma, J; Nakano, I; Napier, A; Necula, V; Neu, C; Neubauer, M S; Nielsen, J; Nigmanov, T; Nodulman, L; Norniella, O; Nurse, E; Oh, S H; Oh, Y D; Oksuzian, I; Okusawa, T; Oldeman, R; Orava, R; Osterberg, K; Pagliarone, C; Palencia, E; Papadimitriou, V; Paramonov, A A; Parks, B; Pashapour, S; Patrick, J; Pauletta, G; Paulini, M; Paus, C; Pellett, D E; Penzo, A; Phillips, T J; Piacentino, G; Piedra, J; Pinera, L; Pitts, K; Plager, C; Pondrom, L; Portell, X; Poukhov, O; Pounder, N; Prakoshyn, F; Pronko, A; Proudfoot, J; Ptohos, F; Punzi, G; Pursley, J; Rademacker, J; Rahaman, A; Ranjan, N; Rappoccio, S; Reisert, B; Rekovic, V; Renton, P; Rescigno, M; Richter, S; Rimondi, F; Ristori, L; Robson, A; Rodrigo, T; Rogers, E; Rolli, S; Roser, R; Rossi, M; Rossin, R; Ruiz, A; Russ, J; Rusu, V; Saarikko, H; Sabik, S; Safonov, A; Sakumoto, W K; Salamanna, G; Saltó, O; Saltzberg, D; Sánchez, C; Santi, L; Sarkar, S; Sartori, L; Sato, K; Savard, P; Savoy-Navarro, A; Scheidle, T; Schlabach, P; Schmidt, E E; Schmidt, M P; Schmitt, M; Schwarz, T; Scodellaro, L; Scott, A L; Scribano, A; Scuri, F; Sedov, A; Seidel, S; Seiya, Y; Semenov, A; Sexton-Kennedy, L; Sfyrla, A; Shapiro, M D; Shears, T; Shepard, P F; Sherman, D; Shimojima, M; Shochet, M; Shon, Y; Shreyber, I; Sidoti, A; Sinervo, P; Sisakyan, A; Sjolin, J; Slaughter, A J; Slaunwhite, J; Sliwa, K; Smith, J R; Snider, F D; Snihur, R; Soderberg, M; Soha, A; Somalwar, S; Sorin, V; Spalding, J; Spinella, F; Spreitzer, T; Squillacioti, P; Stanitzki, M; Staveris-Polykalas, A; St Denis, R; Stelzer, B; Stelzer-Chilton, O; Stentz, D; Strologas, J; Stuart, D; Suh, J S; Sukhanov, A; Sun, H; Suzuki, T; Taffard, A; Takashima, R; Takeuchi, Y; Takikawa, K; Tanaka, M; Tanaka, R; Tecchio, M; Teng, P K; Terashi, K; Thom, J; Thompson, A S; Thomson, E; Tipton, P; Tiwari, V; Tkaczyk, S; Toback, D; Tokar, S; Tollefson, K; Tomura, T; Tonelli, D; Torre, S; Torretta, D; Tourneur, S; Trischuk, W; Tsuchiya, R; Tsuno, S; Turini, N; Ukegawa, F; Unverhau, T; Uozumi, S; Usynin, D; Vallecorsa, S; van Remortel, N; Varganov, A; Vataga, E; Vázquez, F; Velev, G; Veramendi, G; Veszpremi, V; Vidal, R; Vila, I; Vilar, R; Vine, T; Vollrath, I; Volobouev, I; Volpi, G; Würthwein, F; Wagner, P; Wagner, R G; Wagner, R L; Wagner, J; Wagner, W; Wallny, R; Wang, S M; Warburton, A; Waschke, S; Waters, D; Weinberger, M; Wester, W C; Whitehouse, B; Whiteson, D; Wicklund, A B; Wicklund, E; Williams, G; Williams, H H; Wilson, P; Winer, B L; Wittich, P; Wolbers, S; Wolfe, C; Wright, T; Wu, X; Wynne, S M; Yagil, A; Yamamoto, K; Yamaoka, J; Yamashita, T; Yang, C; Yang, U K; Yang, Y C; Yao, W M; Yeh, G P; Yoh, J; Yorita, K; Yoshida, T; Yu, G B; Yu, I; Yu, S S; Yun, J C; Zanello, L; Zanetti, A; Zaw, I; Zhang, X; Zhou, J; Zucchelli, S

2007-06-08

We measure the ratio of cross section times branching fraction, Rp=sigma chi c2 B(chi c2-->J/psi gamma)/sigma chi c1 B(chi c1-->J/psi gamma), in 1.1 fb(-1) of pp collisions at square root s=1.96 TeV. This measurement covers the kinematic range pT(J/psi)>4.0 GeV/c, |eta(J/psi)1.0 GeV/c. For events due to prompt processes, we find Rp=0.395+/-0.016(stat)+/-0.015(syst). This result represents a significant improvement in precision over previous measurements of prompt chi c1,2 hadro production.

Post-irradiation examination of prototype Al-64 wt% U3Si2 fuel rods from NRU

International Nuclear Information System (INIS)

Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D.

1997-01-01

Three prototype fuel rods containing Al-64 wt% U 3 Si 2 (3.15 gU/cm 3 ) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U 3 Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U 3 Si 2 powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37 degrees C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U 3 Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL's research reactors

Progress report for FACETS (Framework Application for Core-Edge Transport Simulations): C.S. SAP

Energy Technology Data Exchange (ETDEWEB)

Epperly, T W

2008-10-01

The mission of the Computer Science Scientific Application Partnership (C.S. SAP) at LLNL is to develop and apply leading-edge scientific component technology to FACETS software. Contributions from LLNL's fusion energy program staff towards the underlying physics modules are described in a separate report. FACETS uses component technology to combine selectively multiple physics and solver software modules written in different languages by different institutions together in an tightly-integrated, parallel computing framework for Tokamak reactor modeling. In the past fiscal year, the C.S. SAP has focused on two primary tasks: applying Babel to connect UEDGE into the FACETS framework through UEDGE's existing Python interface and developing a next generation componentization strategy for UEDGE which avoids the use of Python. The FACETS project uses Babel to solve its language interoperability challenges. Specific accomplishments for the year include: (1) Refined SIDL interfaces for UEDGE to meet satisfy the standard interfaces required by FACETS for all physics modules. This required consensus building between framework and UEDGE developers. (2) Wrote prototype C++ driver for UEDGE to demonstrate how UEDGE can be called from C++ using Babel. (3) Supported the FACETS project by adding new features to Babel such as release number tagging, porting to new machines, and adding new configuration options. Babel modifications were delivered to FACETS by testing and publishing development snapshots in the projects software repository. (4) Assisted Tech-X Corporation in testing and debugging of a high level build system for the complete FACETS tool chain--the complete list of third-party software libraries that FACETS depends on directly or indirectly (e.g., MPI, HDF5, PACT, etc.). (5) Designed and implemented a new approach to wrapping UEDGE as a FACETS component without requiring Python. To get simulation results as soon as possible, our initial connection from the

Laccase-catalyzed C-S and C-C coupling for a one-pot synthesis of 1,4-naphthoquinone sulfides and 1,4-naphthoquinone sulfide dimers

CSIR Research Space (South Africa)

Wellington, Kevin W

2013-06-01

Full Text Available Oxidative C-S and C-C bond formation with aryl and alkyl thiols was catalyzed under mild conditions in a reaction vessel open to air at pH 4.5 in the presence of a commercial laccase (Novozym 51003 or Suberase) and a cosolvent (DMF) to afford 1...

Design and Beam Test Results for the sPHENIX Electromagnetic and Hadronic Calorimeter Prototypes

Energy Technology Data Exchange (ETDEWEB)

Aidala, C.A.; et al.

2017-04-05

The sPHENIX experiment at the Relativistic Heavy Ion Collider (RHIC) will perform high precision measurements of jets and heavy flavor observables for a wide selection of nuclear collision systems, elucidating the microscopic nature of strongly interacting matter ranging from nucleons to the strongly coupled quark-gluon plasma. A prototype of the sPHENIX calorimeter system was tested at the Fermilab Test Beam Facility as experiment T-1044 in the spring of 2016. The electromagnetic calorimeter (EMCal) prototype is composed of scintillating fibers embedded in a mixture of tungsten powder and epoxy. The hadronic calorimeter (HCal) prototype is composed of tilted steel plates alternating with plastic scintillator. Results of the test beam reveal the energy resolution for electrons in the EMCal is $2.8\\%\\oplus~15.5\\%/\\sqrt{E}$ and the energy resolution for hadrons in the combined EMCal plus HCal system is $13.5\\%\\oplus 64.9\\%/\\sqrt{E}$. These results demonstrate that the performance of the proposed calorimeter system is consistent with \\geant simulations and satisfies the sPHENIX specifications.

Utilization of virtual prototyping in development of CMM

International Nuclear Information System (INIS)

Raneda, A.; Pessi, P.; Siuko, M.; Handroos, H.; Palmer, J.; Vilenius, M.

2003-01-01

The characteristic advantages of hydraulics (high power density, simple construction and reliability) together with the characteristics of water as the pressure medium (fire and environmentally safe, chemically neutral, not activated nor affected by radiation) are highlighted in critical applications such as remote handling operations in International Thermonuclear Experimental Reactor (ITER). However, component cost and lack of wide selection of water hydraulic components make it difficult to build and to test complex water hydraulic systems. The use of virtual prototyping for the development of water hydraulic tools can be used to address this problem. Rapidly increased computational power has created conditions for extensive numerical calculations, enabling computer aided virtual prototyping to replace physical prototype phases in product development

Limits and conditions for continuous operation of WWR-S reactor

International Nuclear Information System (INIS)

Pittermann, P.; Listik, E.

1979-02-01

The fundamental technological and nuclear characteristics of the WWR-S reactor, safety limits and concepts of technical surveillance with particular attention to radiation safety of staff and of neighbouring population are outlined. The rules are mandatory for the reactor staff and for the users. The material is part of safety documentation for the WWR-S reactor. (author)

Examples of CEA managements of spent fuels from a prototype power reactor (PHENIX) and from commercial power reactors after post irradiation examinations

International Nuclear Information System (INIS)

Guay, P.

1988-01-01

CEA gained a good experience in the management of spent fuels from its research or power prototype reactors and of the fuel samples for post irradiation examinations. The solution for these products is the reprocessing. The delay to apply that solution is bound to the disponibility of the reprocessing facilities, and in several cases induce a delayed reprocessing. Only particular and limited fuels are planned to be sent in a definitive storage. The definitive storage is choosen only for a few fuels essentially requiring important modifications of the dissolution process. The treatments and operations on the spent fuels must be carried out following the French safety rules. Long and detailed flowsheet studies are therefore necessary before the setting up of the operations. Generally the cost of the management of limited quantities of fuels, as it is the case here, is high. The flowsheets are established in taking into account, as far as possible, the use of existing facilities, procedures, transport casks

The Steam Generating Heavy Water Reactor

International Nuclear Information System (INIS)

Middleton, J.E.

1975-01-01

An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

Progress report on fast breeder reactor development in Japan

International Nuclear Information System (INIS)

1979-01-01

The experimental fast reactor ''Joyo'' will be tested at 75 MW output, starting in April, 1980. In connection with the accident in the Three Mile Island plant, the reexamination of the plant safety and the rechecking-up of the maintenance control system were carried out, and the special inspection by the Science and Technology Agency was executed from May 21 to 23, 1979. Thereafter, the preparation for raising the power output was completed. The periodical inspection after the completion of 50 MW operation is being carried out. The state of progress of various equipments and the codes for core characteristic analysis is reported. The construction preliminary design (2) of the prototype reactor ''Monju'' is examined, and the same design (3) is prepared. The analysis of the decay heat in the prototype reactor is carried on for the safety licensing. The technological investigation of LMFBRs in foreign countries is under way. The preliminary design (4) of the demonstration reactor is under examination, and the technical specifications of the conceptual design (1) are prepared. The researches and developments of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and materials, safety and steam generators are reported. (Kako, I.)

Modelling and experimental study of low temperature energy storage reactor using cementitious material

International Nuclear Information System (INIS)

Ndiaye, Khadim; Ginestet, Stéphane; Cyr, Martin

2017-01-01

Highlights: • Numerical study of a thermochemical reactor using a cementitious material for TES. • Development and test of an original prototype based on this original material. • Comparison of the experimental and numerical results. • Energy balance of the experimental setup (charging and discharging phases). - Abstract: Renewable energy storage is now essential to enhance the energy performance of buildings and to reduce their environmental impact. Most adsorbent materials are capable of storing heat, in a large range of temperature. Ettringite, the main product of the hydration of sulfoaluminate binders, has the advantage of high energy storage density at low temperature, around 60 °C. The objective of this study is, first, to predict the behaviour of the ettringite based material in a thermochemical reactor during the heat storage process, by heat storage modelling, and then to perform experimental validation by tests on a prototype. A model based on the energy and mass balance in the cementitious material was developed and simulated in MatLab software, and was able to predict the spatiotemporal behaviour of the storage system. This helped to build a thermochemical reactor prototype for heat storage tests in both the charging and discharging phases. Thus experimental tests validated the numerical model and served as proof of concept.

Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

Energy Technology Data Exchange (ETDEWEB)

Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

2017-08-30

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO₃-xPbTiO₃) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.

Advanced Reactor Licensing: Experience with Digital I and C Technology in Evolutionary Plants

International Nuclear Information System (INIS)

Wood, RT

2004-01-01

This report presents the findings from a study of experience with digital instrumentation and controls (I and C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l and C systems and identified lessons learned. The report (1) gives an overview of the modern l and C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States

New measuring and protection system at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Jurickova, M.

2006-01-01

The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

Observation of $B^0_s\\rightarrow\\chi_{c1}\\phi$ decay and study of $B^0\\rightarrow\\chi_{c1,2}K^{*0}$ decays

CERN Document Server

INSPIRE-00258707; Adeva, B; Adinolfi, M; Adrover, C; Affolder, A; Ajaltouni, Z; Albrecht, J; Alessio, F; Alexander, M; Ali, S; Alkhazov, G; Alvarez Cartelle, P; Alves Jr, A A; Amato, S; Amerio, S; Amhis, Y; Anderlini, L; Anderson, J; Andreassen, R; Andrews, J E; Appleby, R B; Aquines Gutierrez, O; Archilli, F; Artamonov, A; Artuso, M; Aslanides, E; Auriemma, G; Baalouch, M; Bachmann, S; Back, J J; Baesso, C; Balagura, V; Baldini, W; Barlow, R J; Barschel, C; Barsuk, S; Barter, W; Bauer, Th; Bay, A; Beddow, J; Bedeschi, F; Bediaga, I; Belogurov, S; Belous, K; Belyaev, I; Ben-Haim, E; Bencivenni, G; Benson, S; Benton, J; Berezhnoy, A; Bernet, R; Bettler, M -O; van Beuzekom, M; Bien, A; Bifani, S; Bird, T; Bizzeti, A; Bjørnstad, P M; Blake, T; Blanc, F; Blouw, J; Blusk, S; Bocci, V; Bondar, A; Bondar, N; Bonivento, W; Borghi, S; Borgia, A; Bowcock, T J V; Bowen, E; Bozzi, C; Brambach, T; van den Brand, J; Bressieux, J; Brett, D; Britsch, M; Britton, T; Brook, N H; Brown, H; Burducea, I; Bursche, A; Busetto, G; Buytaert, J; Cadeddu, S; Callot, O; Calvi, M; Calvo Gomez, M; Camboni, A; Campana, P; Campora Perez, D; Carbone, A; Carboni, G; Cardinale, R; Cardini, A; Carranza-Mejia, H; Carson, L; Carvalho Akiba, K; Casse, G; Castillo Garcia, L; Cattaneo, M; Cauet, Ch; Cenci, R; Charles, M; Charpentier, Ph; Chen, P; Chiapolini, N; Chrzaszcz, M; Ciba, K; Cid Vidal, X; Ciezarek, G; Clarke, P E L; Clemencic, M; Cliff, H V; Closier, J; Coca, C; Coco, V; Cogan, J; Cogneras, E; Collins, P; Comerma-Montells, A; Contu, A; Cook, A; Coombes, M; Coquereau, S; Corti, G; Couturier, B; Cowan, G A; Craik, D C; Cunliffe, S; Currie, R; D'Ambrosio, C; David, P; David, P N Y; Davis, A; De Bonis, I; De Bruyn, K; De Capua, S; De Cian, M; De Miranda, J M; De Paula, L; De Silva, W; De Simone, P; Decamp, D; Deckenhoff, M; Del Buono, L; Déléage, N; Derkach, D; Deschamps, O; Dettori, F; Di Canto, A; Di Ruscio, F; Dijkstra, H; Dogaru, M; Donleavy, S; Dordei, F; Dosil Suárez, A; Dossett, D; Dovbnya, A; Dupertuis, F; Dzhelyadin, R; Dziurda, A; Dzyuba, A; Easo, S; Egede, U; Egorychev, V; Eidelman, S; van Eijk, D; Eisenhardt, S; Eitschberger, U; Ekelhof, R; Eklund, L; El Rifai, I; Elsasser, Ch; Elsby, D; Falabella, A; Färber, C; Fardell, G; Farinelli, C; Farry, S; Fave, V; Ferguson, D; Fernandez Albor, V; Ferreira Rodrigues, F; Ferro-Luzzi, M; Filippov, S; Fiore, M; Fitzpatrick, C; Fontana, M; Fontanelli, F; Forty, R; Francisco, O; Frank, M; Frei, C; Frosini, M; Furcas, S; Furfaro, E; Gallas Torreira, A; Galli, D; Gandelman, M; Gandini, P; Gao, Y; Garofoli, J; Garosi, P; Garra Tico, J; Garrido, L; Gaspar, C; Gauld, R; Gersabeck, E; Gersabeck, M; Gershon, T; Ghez, Ph; Gibson, V; Giubega, L; Gligorov, V V; Göbel, C; Golubkov, D; Golutvin, A; Gomes, A; Gordon, H; Grabalosa Gándara, M; Graciani Diaz, R; Granado Cardoso, L A; Graugés, E; Graziani, G; Grecu, A; Greening, E; Gregson, S; Griffith, P; Grünberg, O; Gui, B; Gushchin, E; Guz, Yu; Gys, T; Hadjivasiliou, C; Haefeli, G; Haen, C; Haines, S C; Hall, S; Hamilton, B; Hampson, T; Hansmann-Menzemer, S; Harnew, N; Harnew, S T; Harrison, J; Hartmann, T; He, J; Head, T; Heijne, V; Hennessy, K; Henrard, P; Hernando Morata, J A; van Herwijnen, E; Hicheur, A; Hicks, E; Hill, D; Hoballah, M; Holtrop, M; Hombach, C; Hopchev, P; Hulsbergen, W; Hunt, P; Huse, T; Hussain, N; Hutchcroft, D; Hynds, D; Iakovenko, V; Idzik, M; Ilten, P; Jacobsson, R; Jaeger, A; Jans, E; Jaton, P; Jawahery, A; Jing, F; John, M; Johnson, D; Jones, C R; Joram, C; Jost, B; Kaballo, M; Kandybei, S; Kanso, W; Karacson, M; Karbach, T M; Kenyon, I R; Ketel, T; Keune, A; Khanji, B; Kochebina, O; Komarov, I; Koopman, R F; Koppenburg, P; Korolev, M; Kozlinskiy, A; Kravchuk, L; Kreplin, K; Kreps, M; Krocker, G; Krokovny, P; Kruse, F; Kucharczyk, M; Kudryavtsev, V; Kvaratskheliya, T; La Thi, V N; Lacarrere, D; Lafferty, G; Lai, A; Lambert, D; Lambert, R W; Lanciotti, E; Lanfranchi, G; Langenbruch, C; Latham, T; Lazzeroni, C; Le Gac, R; van Leerdam, J; Lees, J -P; Lefèvre, R; Leflat, A; Lefrançois, J; Leo, S; Leroy, O; Lesiak, T; Leverington, B; Li, Y; Li Gioi, L; Liles, M; Lindner, R; Linn, C; Liu, B; Liu, G; Lohn, S; Longstaff, I; Lopes, J H; Lopez-March, N; Lu, H; Lucchesi, D; Luisier, J; Luo, H; Machefert, F; Machikhiliyan, I V; Maciuc, F; Maev, O; Malde, S; Manca, G; Mancinelli, G; Marconi, U; Märki, R; Marks, J; Martellotti, G; Martens, A; Martín Sánchez, A; Martinelli, M; Martinez Santos, D; Martins Tostes, D; Massafferri, A; Matev, R; Mathe, Z; Matteuzzi, C; Maurice, E; Mazurov, A; Mc Skelly, B; McCarthy, J; McNab, A; McNulty, R; Meadows, B; Meier, F; Meissner, M; Merk, M; Milanes, D A; Minard, M -N; Molina Rodriguez, J; Monteil, S; Moran, D; Morawski, P; Mordà, A; Morello, M J; Mountain, R; Mous, I; Muheim, F; Müller, K; Muresan, R; Muryn, B; Muster, B; Naik, P; Nakada, T; Nandakumar, R; Nasteva, I; Needham, M; Neubert, S; Neufeld, N; Nguyen, A D; Nguyen, T D; Nguyen-Mau, C; Nicol, M; Niess, V; Niet, R; Nikitin, N; Nikodem, T; Nomerotski, A; Novoselov, A; Oblakowska-Mucha, A; Obraztsov, V; Oggero, S; Ogilvy, S; Okhrimenko, O; Oldeman, R; Orlandea, M; Otalora Goicochea, J M; Owen, P; Oyanguren, A; Pal, B K; Palano, A; Palutan, M; Panman, J; Papanestis, A; Pappagallo, M; Parkes, C; Parkinson, C J; Passaleva, G; Patel, G D; Patel, M; Patrick, G N; Patrignani, C; Pavel-Nicorescu, C; Pazos Alvarez, A; Pellegrino, A; Penso, G; Pepe Altarelli, M; Perazzini, S; Perez Trigo, E; Pérez-Calero Yzquierdo, A; Perret, P; Perrin-Terrin, M; Pessina, G; Petridis, K; Petrolini, A; Phan, A; Picatoste Olloqui, E; Pietrzyk, B; Pilař, T; Pinci, D; Playfer, S; Plo Casasus, M; Polci, F; Polok, G; Poluektov, A; Polyakov, I; Polycarpo, E; Popov, A; Popov, D; Popovici, B; Potterat, C; Powell, A; Prisciandaro, J; Pritchard, A; Prouve, C; Pugatch, V; Puig Navarro, A; Punzi, G; Qian, W; Rademacker, J H; Rakotomiaramanana, B; Rangel, M S; Raniuk, I; Rauschmayr, N; Raven, G; Redford, S; Reid, M M; dos Reis, A C; Ricciardi, S; Richards, A; Rinnert, K; Rives Molina, V; Roa Romero, D A; Robbe, P; Rodrigues, E; Rodriguez Perez, P; Roiser, S; Romanovsky, V; Romero Vidal, A; Rouvinet, J; Ruf, T; Ruffini, F; Ruiz, H; Ruiz Valls, P; Sabatino, G; Saborido Silva, J J; Sagidova, N; Sail, P; Saitta, B; Salustino Guimaraes, V; Salzmann, C; Sanmartin Sedes, B; Sannino, M; Santacesaria, R; Santamarina Rios, C; Santovetti, E; Sapunov, M; Sarti, A; Satriano, C; Satta, A; Savrie, M; Savrina, D; Schaack, P; Schiller, M; Schindler, H; Schlupp, M; Schmelling, M; Schmidt, B; Schneider, O; Schopper, A; Schune, M -H; Schwemmer, R; Sciascia, B; Sciubba, A; Seco, M; Semennikov, A; Sepp, I; Serra, N; Serrano, J; Seyfert, P; Shapkin, M; Shapoval, I; Shatalov, P; Shcheglov, Y; Shears, T; Shekhtman, L; Shevchenko, O; Shevchenko, V; Shires, A; Silva Coutinho, R; Sirendi, M; Skwarnicki, T; Smith, N A; Smith, E; Smith, J; Smith, M; Sokoloff, M D; Soler, F J P; Soomro, F; Souza, D; Souza De Paula, B; Spaan, B; Sparkes, A; Spradlin, P; Stagni, F; Stahl, S; Steinkamp, O; Stoica, S; Stone, S; Storaci, B; Straticiuc, M; Straumann, U; Subbiah, V K; Sun, L; Swientek, S; Syropoulos, V; Szczekowski, M; Szczypka, P; Szumlak, T; T'Jampens, S; Teklishyn, M; Teodorescu, E; Teubert, F; Thomas, C; Thomas, E; van Tilburg, J; Tisserand, V; Tobin, M; Tolk, S; Tonelli, D; Topp-Joergensen, S; Torr, N; Tournefier, E; Tourneur, S; Tran, M T; Tresch, M; Tsaregorodtsev, A; Tsopelas, P; Tuning, N; Ubeda Garcia, M; Ukleja, A; Urner, D; Ustyuzhanin, A; Uwer, U; Vagnoni, V; Valenti, G; Vallier, A; Van Dijk, M; Vazquez Gomez, R; Vazquez Regueiro, P; Vázquez Sierra, C; Vecchi, S; Velthuis, J J; Veltri, M; Veneziano, G; Vesterinen, M; Viaud, B; Vieira, D; Vilasis-Cardona, X; Vollhardt, A; Volyanskyy, D; Voong, D; Vorobyev, A; Vorobyev, V; Voß, C; Voss, H; Waldi, R; Wallace, C; Wallace, R; Wandernoth, S; Wang, J; Ward, D R; Watson, N K; Webber, A D; Websdale, D; Whitehead, M; Wicht, J; Wiechczynski, J; Wiedner, D; Wiggers, L; Wilkinson, G; Williams, M P; Williams, M; Wilson, F F; Wimberley, J; Wishahi, J; Witek, M; Wotton, S A; Wright, S; Wu, S; Wyllie, K; Xie, Y; Xing, Z; Yang, Z; Young, R; Yuan, X; Yushchenko, O; Zangoli, M; Zavertyaev, M; Zhang, F; Zhang, L; Zhang, W C; Zhang, Y; Zhelezov, A; Zhokhov, A; Zhong, L; Zvyagin, A

2013-01-01

The first observation of the decay $B^0_s\\rightarrow\\chi_{c1}\\phi$ and a study of $B^0\\rightarrow\\chi_{c1,2}K^{*0}$ decays are presented. The analysis is performed using a dataset, corresponding to an integrated luminosity of 1.0 fb$^{-1}$, collected by the LHCb experiment in pp collisions at a centre-of-mass energy of 7 TeV. The following ratios of branching fractions are measured: \\begin{equation*} \\begin{array}{lll} \\dfrac{\\cal{B}(B^0_s\\rightarrow\\chi_{c1}\\phi)}{\\cal{B}(B^0_s\\rightarrow J/\\psi\\phi)} &=& (18.9 \\pm1.8\\,(stat)\\pm1.3\\,(syst)\\pm0.8\\,(\\cal{B})) \\times 10^{-2}, \\\\ \\dfrac{\\cal{B}(B^0\\rightarrow\\chi_{c1}K^{*0})}{\\cal{B}(B^0\\rightarrow J/\\psi K^{*0})} &=& (19.8 \\pm1.1\\,(stat)\\pm1.2\\,(syst)\\pm0.9\\,(\\cal{B})) \\times 10^{-2}, \\\\ \\dfrac{\\cal{B}(B^0\\rightarrow\\chi_{c2}K^{*0})}{\\cal{B}(B^0\\rightarrow\\chi_{c 1}K^{*0})} &=& (17.1 \\pm5.0\\,(stat)\\pm1.7\\,(syst)\\pm1.1\\,(\\cal{B})) \\times 10^{-2}, \\\\ \\end{array} \\end{equation*} where the third uncertainty is due to the limited knowledge o...

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

Science.gov (United States)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

Halden reactor project

International Nuclear Information System (INIS)

1980-01-01

The research programme at the Halden Project is focused on the following three areas: 1. In-core behavior of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2. Prediction, surveillance and control of fuel and core performance for which models of fuel and core behavior are developed. 3. Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

Energy Technology Data Exchange (ETDEWEB)

Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

1997-08-01

Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

Evaluación del comportamiento hidráulico en un reactor anaerobio de doble cámara (RADCA

Directory of Open Access Journals (Sweden)

Nancy Rincón

2011-01-01

tales como cortos circuitos, zonas muertas y recirculación interna afectan su desempeño. En esta investigación se evaluó el comportamiento hidráulico de un reactor anaerobio de doble cámara (RADCA de 534,5 L (cámara 1=305 L y cámara 2= 229,5 L como innovación tecnológica de los reactores UASB. El RADCA fue alimentado con agua residual municipal (ARM de la ciudad de Maracaibo, Venezuela; cada una de las cámaras fueron inoculadas con lodo granular (20% v/v proveniente de una cervecería local. La evaluación hidráulica se realizó en la fase líquida y en operación utilizando Li+ (LiCl como trazador aplicado de forma instantánea en el afluente a tiempo de retención hidráulico teórico (TRHt de 6 horas; 3,4 h en la cámara 1 y 2,6 para la cámara 2. El RADCA describió un flujo pistón en ambas cámaras y una eficiencia hidráulica cercana a la unidad (1 indicando una presencia casi nula de zonas muertas. La eficiencia de remoción de la DQO total (DQOT del RADCA se mantuvo en el rango de 59,77% a 74,64% con un promedio de 68,26%. Para las cámaras 1 y 2 la eficiencia promedio fue 60,4 y 20,94% con una producción de biogás (L/h de 2,768 y 0,541 respectivamente.

Prototypic Enhanced Risk Monitor Framework and Evaluation - Advanced Reactor Technology Milestone: M3AT-15PN2301054

Energy Technology Data Exchange (ETDEWEB)

Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Veeramany, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bonebrake, Christopher A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ivans, William J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coles, Garill A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coble, Jamie B. [Univ. of Tennessee, Knoxville, TN (United States); Liu, X. [Univ. of Tennessee, Knoxville, TN (United States); Wootan, David W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mitchell, Mark R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Brass, Mary F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-09-24

This research report summaries the development and evaluation of a prototypic enhanced risk monitor (ERM) methodology (framework) that includes alternative risk metrics and uncertainty analysis. This updated ERM methodology accounts for uncertainty in the equipment condition assessment (ECA), the prognostic result, and the probabilistic risk assessment (PRA) model. It is anticipated that the ability to characterize uncertainty in the estimated risk and update the risk estimates in real time based on equipment condition assessment (ECA) will provide a mechanism for optimizing plant performance while staying within specified safety margins. These results (based on impacting active component O&M using real-time equipment condition information) are a step towards ERMs that, if integrated with AR supervisory plant control systems, can help control O&M costs and improve affordability of advanced reactors.

Overview of fast reactor safety research and development in the USA

International Nuclear Information System (INIS)

Neuhold, R.J.; Avery, R.; Marchaterre, J.F.

1986-01-01

The liquid metal reactor (LMR) safety R and D program in the U.S. is presently focused on support of two modular innovative reactor concepts: PRISM - the General Electric Power Reactor Inherently Safe Module and SAFR - the Rockwell International Sodium Advanced Fast Reactor. These reactor plant concepts accommodate the use of either oxide fuel or the metal fuel which is under development in the Argonne National Laboratory (ANL) Integral Fast Reactor (IFR) program. Both concepts emphasize prevention of accidents through enhancement of inherent and passive safety characteristics. Enhancement of these characteristics is expected to be a major factor in establishing new and improved safety criteria and licensing arrangements with regulatory authorities for advanced reactors. Limited work is also continuing on the Large Scale Prototype Breeder (LSPB), a large pool plant design. Major elements of the current and restructured safety program are discussed. (author)

Controls and interlocks for a prototype 1MHz beam chopper

International Nuclear Information System (INIS)

Waters, G.; Bishop, D.; Barnes, M.J.; Wait, G.D.

1991-05-01

A prototype 1 MHz beam chopper for the proposed KAON Factory at TRIUMF has been constructed. The chopper is an electric field device, driven by a tetrode based pulser, for deflecting a charged particle beam. Associated with the tetrode used in the prototype design are high voltage power supplies for the electrodes. We use an FET based grid pulser and a sequencer capable of accurate digital control of pulse timing to 0.4 ns. A safety interlock and control system using a programmable controller with fibre optic links has been built. This has given us the versatility required in a prototype system. (Author) 4 refs., 5 figs

$\\chi^{\\vphantom\\dagger}_{c0}(3915)$ As the Lightest $c\\bar c s \\bar s$ State

CERN Document Server

Lebed, Richard F.

2016-05-23

The state $\\chi^{\\vphantom\\dagger}_{c0}(3915)$ has recently been demoted by the Particle Data Group from its previous status as the conventional $c\\bar c$ $2 {}^3P_0$ state, largely due to the absence of expected $D\\bar D$ decays. We propose that $\\chi^{\\vphantom\\dagger}_{c0}(3915)$ is actually the lightest $c\\bar c s \\bar s$ state, and calculate the spectrum of such states using the diquark model, identifying many of the observed charmoniumlike states that lack open-charm decay modes as $c\\bar c s \\bar s$. Among other results, we argue that $Y(4140)$ is a $J^{PC} = 1^{++}$ $c\\bar c s \\bar s$ state that has been not been seen in two-photon fusion largely as a consequence of the Landau-Yang theorem.

Wet storage of nuclear spent fuel from nuclear research reactor WWR-S

International Nuclear Information System (INIS)

Dragolici, A. C; Zorliu, A.; Petran, C.; Mincu, I.

2001-01-01

robustness of this type of repository. The ponds are supplied with distilled water from reactor tank No 1. The spent fuel assemblies used in the WWR-S reactor and stored in this repository contain two types of fuel: the EK-10 (153 pieces) and the S-36 (70 pieces). The EK-10 and S-36 fuel assemblies have the same shape but distinct numbers of fuel rods and different percentage of U-235 enrichment. So, the EK-10 contains 16 fuel rods and 10% enrichment while the S-36 contains 15 fuel rods of 36% U-235 enrichment. Total activity of the fuel assemblies is about 10.69 x 10 4 Ci. The most important fission products in fuel rods are Sr-90 and Cs-137. Depending on fuel assembly type the calculated values of the activity of these elements are: - a. 7.14 x 10 3 Ci (Sr-90) and 7.67 x 10 3 Ci (Cs-137) for EK-10 and, - b. 8.21 x 10 3 Ci (Sr-90) and 8.08 x 10 3 Ci (Cs-137) for S-36. The gamma spectrometry analysis performed on water samples from storage ponds showed the dominant presence of Cs and other nuclides. The presence of Cs in the water is due to fuel rods flaws, the occurrence of pores and cracks because of fuel manipulation in active core of the reactor or in the storage pond or because of corrosion pits due to a low quality of water. The manufacturer have specified the water conditions for the EK-10 and S-36 fuel assemblies, which they believe will avoid corrosion problems. These are displayed in a table which enlists for EK-10 and C-36 the values of the water parameters, namely the values of pH, conductivity, constant residuals, corrosion products, Cl - and O 2 concentration. The IAEA recommends values of 1 μS/cm, but values up to 2 μS/cm might be accepted if the chlorine, copper and sulphate concentrations are very low. To maintain these parameters of the water in compliance with IAEA recommendations, it was designed and realised a filtration installation for the distilled water produced at the reactor. Till this installation will be assembled the water chemistry is

Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

Energy Technology Data Exchange (ETDEWEB)

1994-01-15

This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports

International Nuclear Information System (INIS)

1994-01-01

This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A ampersand 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met

Development of a Test Facility to Simulate the Reactor Flow Distribution of APR+

International Nuclear Information System (INIS)

Euh, D. J.; Cho, S.; Youn, Y. J.; Kim, J. T.; Kang, H. S.; Kwon, T. S.

2011-01-01

Recently a design of new reactor, APR+, is being developed, as an advanced type of APR1400. In order to analyze the thermal margin and hydraulic characteristics of APR+, quantification tests for flow and pressure distribution with a conservation of flow geometry are necessary. Hetsroni (1967) proposed four principal parameters for a hydraulic model representing a nuclear reactor prototype: geometry, relative roughness, Reynolds number, and Euler number. He concluded that the Euler number should be similar in the prototype and model under the preservation of the aspect ratio on the flow path. The effect of the Reynolds number at its higher values on the Euler number is rather small, since the dependency of the form and frictional loss coefficients on the Reynolds number is seen to be small. ABB-CE has carried out several reactor flow model test programs, mostly for its prototype reactors. A series of tests were conducted using a 3/16 scale reactor model. (see Lee et al., 2001). Lee et al (1991) performed experimental studies using a 1/5.03 scale reactor flow model of Yonggwang nuclear units 3 and 4. They showed that the measured data met the acceptance criteria and were suitable for their intended use in terms of performance and safety analyses. The design of current test facility was based on the conservation of Euler number which is a ratio of pressure drop to dynamic pressure with a sufficiently turbulent region having a high Reynolds number. By referring to the previous study, the APR+ design is linearly reduced to 1/5 ratio with a 1/2 of the velocity scale, which yields a 1/39.7 of Reynolds number scaling ratio. In the present study, the design feature of the facilities, named 'ACOP', in order to investigate flow and pressure distribution are described

Development and applications of reactor noise analysis at Ontario Hydro`s CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O [Ontario Hydro, Toronto, ON (Canada); Tulett, M V [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro`s CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

Energy Technology Data Exchange (ETDEWEB)

Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

2017-05-15

Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

Novel infectious cDNA clones of hepatitis C virus genotype 3a (strain S52) and 4a (strain ED43): genetic analyses and in vivo pathogenesis studies

DEFF Research Database (Denmark)

Gottwein, Judith; Scheel, Troels; Callendret, Benoit

2010-01-01

Previously, RNA transcripts of cDNA clones of hepatitis C virus (HCV) genotypes 1a (strains H77, HCV-1, and HC-TN), 1b (HC-J4, Con1, and HCV-N), and 2a (HC-J6 and JFH1) were found to be infectious in chimpanzees. However, only JFH1 was infectious in human hepatoma Huh7 cells. We performed genetic...... analysis of HCV genotype 3a (strain S52) and 4a (strain ED43) prototype strains and generated full-length consensus cDNA clones (pS52 and pED43). Transfection of Huh7.5 cells with RNA transcripts of these clones did not yield cells expressing HCV Core. However, intrahepatic transfection of chimpanzees...... resulted in robust infection with peak HCV RNA titers of approximately 5.5 log(10) international units (IU)/ml. Genomic consensus sequences recovered from serum at the times of peak viral titers were identical to the sequences of the parental plasmids. Both chimpanzees developed acute hepatitis...

Liquid metal reactor head designs in the USA - heat and mass transfer considerations

International Nuclear Information System (INIS)

Burke, T.M.

1986-01-01

Development of liquid metal reactor plants in the United States over the past 30 years has resulted in an evolution of reactor head designs as reflected in the SRE, Hallam, EBR-II and FFTF plants. This evolution has probably been affected to some extent by the fact that, in contrast to most other countries, there is no single organization in the United States which has been responsible for the design of liquid metal reactor plants. The current U.S. LMR design efforts involve two innovative design consortiums (guided by the US Department of Energy) and a joint industry venture on the Large Scale Prototype Breeder. It is therefore somewhat difficult to provide a statement on the philosophy of the reactor head design in the U.S. This paper however briefly describes the existing and proposed U.S. liquid metal reactor head designs and in the process, attempt to provide some insight on the basis for those designs

Verification Test of Hydraulic Performance for Reactor Coolant Pump

Energy Technology Data Exchange (ETDEWEB)

Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

2010-01-15

According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

Creep properties of Nb-1Zr and Nb-1Zr-0.1C

International Nuclear Information System (INIS)

Horak, J.A.; Egner, L.K.

1994-12-01

In the early 1980s a compact, lithium cooled, fast-energy spectrum nuclear reactor was selected for space applications requiring prolonged uninterrupted electrical power. This reactor was to be capable of generating up to 100 kilowatts of electricity for times up to seven years in space and thus was given the acronym SP-100. The material selected for the fuel cladding, reactor heat transport systems and structural components was Nb-1 wt % Zr (Nb-1Zr). In addition to commercial Nb-1Zr, modified alloys containing 100--200 wt ppM each of carbon and nitrogen and 900 ± 150 wt ppM carbon were also included, Type B Nb-1Zr and PWC-11, respectively. The SP-100 reactor was designed to operate at temperatures of 1290--1425 K. At these temperatures the principal mode of deformation for Nb-1Zr is creep, and creep strain of the fuel cladding limits the useful reactor lifetime. To develop a creep data base for design, safety and reliability analyses, uniaxial creep testing of Nb-1Zr, Type B Nb-1Zr and PWC-11 was conducted from 1250--1450 K at stresses from 5.0 MPa to 41.4 MPa. Methodology and test results are presented

Upper limits for the circular dichroism for the C 1s and O 1s core excitation of methyl oxirane

International Nuclear Information System (INIS)

Pruemper, G; Lischke, T; Fukuzawa, H; Reinkoester, A; Ueda, K

2007-01-01

The circular dichroism (CD) in the total and partial ion yields of methyl-oxirane C 3 H 6 O was measured at the C 1s and O 1s edges. The difference of the response of the chiral molecule to circularly polarized light with opposite handedness was found to be less than 0.2% for the total ion yield and less than 0.5% for the partial ion yield. Additionally we tried to find a dipole allowed molecular orientation CD effect by analysing the fragmentation in the forward and backward direction. For this effect we found an upper limit of 1-2% for all abundant ionic fragments

UA1 prototype detector

CERN Multimedia

1980-01-01

Prototype of UA1 central detector inside a plexi tube. The UA1 experiment ran at CERN's Super Proton Synchrotron and made the Nobel Prize winning discovery of W and Z particles in 1983. The UA1 central detector was crucial to understanding the complex topology of proton-antiproton events. It played a most important role in identifying a handful of Ws and Zs among billions of collisions. The detector was essentially a wire chamber - a 6-chamber cylindrical assembly 5.8 m long and 2.3 m in diameter, the largest imaging drift chamber of its day. It recorded the tracks of charged particles curving in a 0.7 Tesla magnetic field, measuring their momentum, the sign of their electric charge and their rate of energy loss (dE/dx). Atoms in the argon-ethane gas mixture filling the chambers were ionised by the passage of charged particles. The electrons which were released drifted along an electric field shaped by field wires and were collected on sense wires. The geometrical arrangement of the 17000 field wires and 6...

Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

Energy Technology Data Exchange (ETDEWEB)