Development of FBR technology in the FBR 'Joyo'

International Nuclear Information System (INIS)

Nara, Yoshihiko; Akiyama, Takao; Sato, Isao; Mizoo, Nobutatsu; Yoshimi, Hirotaka; Shimada, Takashi

1986-01-01

Power Reactor and Nuclear Fuel Development Corp. has advanced the construction of the prototype FBR ''Monju'', and the ground breaking ceremony was held on October 28, 1985. For the design and construction of Monju, the experience, achievement, and the results of development by the own effort and international cooperation gained by the experimental FBR ''Joyo'' have been reflected. It is important to develop the core management technology, operation-supporting system, the techniques of regular inspection, maintenance and repair, the reduction of radiation exposure and so on, to accumulate the experience, and to reflect those accurately to Monju. The operation history of the experimental FBR ''Joyo'', the international joint research on FBRs using the Joyo, the results regarding the characteristic technology of FBRs such as the reactor core, fuel and control rods, sodium technology, the construction of machinery and equipment, and the plant system the plan of developing the high grade technology of FBRs such as the development of fuel and materials, the improvement of reliability and the development of operation management techniques, the verifying test of new technology such as spent fuel storage, the new system for sodium purification and the techniques for analyzing earthquake response, and the international cooperation are reported. (Kako, I.)

Development of wire wrapping technology for FBR fuel pin

International Nuclear Information System (INIS)

Nogami, Tetsuya; Seki, Nobuo; Sawayama, Takeo; Ishibashi, Takashi

1991-01-01

For the FBR fuel assembly, the spacer wire is adopted to maintain the space between fuel pins. The developments have been carried out to achieve automatically wire wrapping with high precision. Based on the fundamental technology developed through the mock-up test operation, Joyo 'MK-I', fuel pin fabrication was started using partially mechanized wire wrapping machine in 1973. In 1978, an automated wire wrapping machine for Joyo 'MK-II' was developed by the adoption of some improvements for the wire inserting system to end plug hole and the precision of wire pitch. On the bases of these experiences, fully automated wire wrapping machine for 'Monju' fuel pin was installed at Plutonium Fuel Production Facility (PFPF) in 1987. (author)

Fast reactor core design studies to cope with TRU fuel composition changes in the LWR-to-FBR transition period

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

2009-01-01

As part of the Fast Reactor Cycle Technology Development Project (FaCT Project), sodium-cooled fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. A 750 MWe mixed-oxide fuel core is firstly defined as a FaCT medium-size reference core and its neutronics characteristics are determined. The core is a high internal conversion type and has an average burnup of 150 GWD/T. The reference TRU fuel composition is assumed to come from the FBR equilibrium state. Compared to the LWR-to-FBR transition period, the TRU fuels in the FBR equilibrium period are multi-recycled through fast reactors and have a different composition. An available TRU fuel composition is determined by fast reactor spent fuel multi-recycling scenarios. Then the FaCT core corresponding to the TRU fuel with different compositions is set according to the TRU fuel composition changes in LWR-to-FBR transition period, and the key core neutronics characteristics are assessed. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters. (author)

Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

International Nuclear Information System (INIS)

Permana, Sidik; Suzuki, Mitsutoshi; Saito, Masaki; Novitrian,; Waris, Abdul; Suud, Zaki

2013-01-01

Highlights: • The paper analyzes the plutonium production of recycling nuclear fuel option. • To evaluate material attractiveness based on intrinsic feature of material barrier. • Evaluation based on isotopic plutonium composition of spent fuel LWR and FBR. • Even mass number of plutonium gives a significant contribution to material barrier, in particular Pu-238 and Pu-240. • Doping MA in FBR blanket is effective to increase material barrier from weapon grade plutonium to more than MOX fuel grade. - Abstract: Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no

TRU composition changes and their influence on FBR core characteristics in the LWR-to-FBR transition period

International Nuclear Information System (INIS)

Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

2009-01-01

In the conceptual core and fuel design studies in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan, much interest has been taken in the fuel nuclide compositions for a transition period from light water reactor to fast breeder reactor (FBR). In this paper, the range of transuranic (TRU) nuclide composition to be provided to FBR is evaluated with extended recycling scenario calculations. The influence of TRU composition changes on FBR core characteristics are also discussed with explanations of major contributing factors. (author)

Sensitivity analysis of fuel pin failure performance under slow-ramp type transient overpower condition by using a fuel performance analysis code FEMAXI-FBR

International Nuclear Information System (INIS)

Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

2012-01-01

The FEMAXI-FBR is a fuel performance analysis code and has been developed as one module of core disruptive evaluation system, the ASTERIA-FBR. The FEMAXI-FBR has reproduced the failure pin behavior during slow transient overpower. The axial location of pin failure affects the power and reactivity behavior during core disruptive accident, and failure model of which pin failure occurs at upper part of pin is used by reflecting the results of the CABRI-2 test. By using the FEMAXI-FBR, sensitivity analysis of uncertainty of design parameters such as irradiation conditions and fuel fabrication tolerances was performed to clarify the effect on axial location of pin failure during slow transient overpower. The sensitivity analysis showed that the uncertainty of design parameters does not affect the failure location. It suggests that the failure model with which locations of failure occur at upper part of pin can be adopted for core disruptive calculation by taking into consideration of design uncertainties. (author)

Design study on metal fuel FBR cores

International Nuclear Information System (INIS)

Yokoo, T.; Tanaka, Y.; Ogata, T.

1991-01-01

A design approach for metal fuel FBR core to maintain fuel integrity during transient events by limiting eutectic/liquid phase formation is proposed based on the current status of metallic fuel development. Its impact as the limitation on the core outlet temperature is assessed through its application to two of CRIEPI's core concepts, high linear power 1000 MWe homogeneous design and medium linear power 300 MWe radially heterogeneous design. SESAME/SALT code is used in this study to analyze steady state and transient fuel behavior. SE2-FA code is developed based on SUPERENERGY-2 and used to analyze core thermal-hydraulics with uncertainties. As the result, the core outlet temperatures of both designs are found to be limited to ≤500degC if it is required to prevent eutectic/liquid phase formation during operational transients in order to guarantee the fuel integrity. Additional assessment is made assuming an advanced limiting condition that allows small liquid phase formation based on the liquid phase penetration rate derived from existing experimental results. The result indicates possibility of raising core outlet temperature to ∼ 530degC. Also, it is found that core design technology improvements such as hot spot factors reduction can contribute to the core outlet temperature extension by 10 ∼ 20degC. (author)

Dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Sanyoshi, H.; Nishina, H.; Toyota, O.; Yamamoto, R.; Nemoto, S.; Okamoto, F.; Togashi, A.; Kawata, T.; Hayashi, S.

1991-01-01

At the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), the Chemical Processing Facility (CPF) has been continuing operation since 1982 for laboratory scale hot experiments on reprocessing of FBR mixed oxide fuel. As a part of these experiments, dissolution experiments have been performed to define the key parameters affecting dissolution rates such as concentration of nitric acid, temperature and burnup and also to confirm the amount of insoluble residue. The dissolution rate of the irradiated fuel was determined to be in proportion to the 1.7 power of the nitric acid concentration. The activation energy determined from the experiments varied from 6 to 11 kcal/mol depending on the method of dissolution. The dissolution rate decreased as the fuel burnup increased in low nitric acid media below 5 mol/l. However, it was found that the effect of the burnup became negligible in a high concentration of nitric acid media. The amount of insoluble residue and its constituents were evaluated by changing the dissolution condition. (author)

Development of simulation code for FBR spent fuel dissolution with rotary drum type continuous dissolver

International Nuclear Information System (INIS)

Sano, Yuichi; Katsurai, Kiyomichi; Washiya, Tadahiro; Koizumi, Tsutomu; Matsumoto, Satoshi

2011-01-01

Japan Atomic Energy Agency (JAEA) has been studying rotary drum type continuous dissolver for FBR spent fuel dissolution. For estimating the fuel dissolution behavior under several operational conditions in this dissolver, we have been developing the simulation code, PLUM, which mainly consists of 3 modules for calculating chemical reaction, mass transfer and thermal balance in the rotary drum type continuous dissolver. Under the various conditions where dissolution experiments were carried out with the batch-wise dissolver for FBR spent fuel and with the rotary drum type continuous dissolver for UO 2 fuel, it was confirmed that the fuel dissolution behaviors calculated by the PLUM code showed good agreement with the experimental ones. Based on this result, the condition for obtaining the dissolver solution with high HM (heavy metal : U and Pu) concentration (∼500g/L), which is required for the next step, i.e. crystallization process, was also analyzed by this code and appropriate operational conditions with the rotary drum type continuous dissolver, such as feedrate, concentration and temperature of nitric acid, could be clarified. (author)

Research on CDA for advanced fuel FBR

International Nuclear Information System (INIS)

Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime.

1997-03-01

For the purpose of evaluating possibility of the re-criticality of a metallic fueled reactor, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'Research on CDA for advanced fuel FBR'. The results of this year are the following. The accident initiator considered is a loss-of-flow accident with ATWS. The LOF analysis was performed for the metallic fueled 600 MWe homogeneous two region reactors, both for a metallic fuel only and for a metallic fuel core with ZrH pin. The SAS3D CDA initiation phase analysis code was used to investigate the re-criticality potential at the severe accident. The change mainly in the constants was necessary to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. LOF with flow decay half time of t 1/2 =0.5(s) (all blackout case) and 5.5(s) (ordinary LOF case) were analyzed. Independent of the conditions of the analysis, the results show all the cases could avoid to become prompt-critical. Depending on the analysis condition, it becomes necessary to transfer to the transient phase, it is also shown there is a possibility to avoid re-criticality due to the motion of molten fuel both for the metallic fuel and for the metallic fuel with ZrH moderator. However, because of the constants used for the material property the results might overestimate the fuel motion. It is shown that the moderator is effective to terminate the accident at an early stage. The behavior of metallic fueled reactors at CDA was analyzed with SAS3D code by modifying the constants of material properties to be applied to the reactor. It is shown that a metallic fueled reactor has a possibility to avoid re-criticality at CDA. (J.P.N.)

Feasibility study of application of ductless fuel assembly to FBR

International Nuclear Information System (INIS)

Itoh, K.; Shibahara, I.

1996-01-01

Feasibility studies on an application of the ductless fuel concept to an FBR core have been carried out in order to evaluate the basic features of the ductless core, especially in the fields of the thermal-hydraulic aspects and the mechanical behaviors. Regarding thermal-hydraulic aspects, analyses were performed by using a whole core thermal-hydraulic analysis code by making some modification for this study on the PWR code THINC. A small scaled ductless core model was prepared and a hydraulic experiment was carried out to study basic hydraulic characteristics of a ductless core. Core mechanical behaviors were analyzed focusing on the core irradiation bowing aspects and the seismic behaviors. Following features are revealed on the core structural behaviors: (1) the bowing stiffness of the ductless assembly is around 1/5 to 1/10 of that of the duct type assembly; (2) the contact loads between assemblies by the bowing effects are small through core cycles; (3) the damping of the ductless assemblies are so large that the seismic responses are small and the loads between assemblies are small due to occurring many contact points. Through this study it is expected that the concept of the ductless fuel can be applicable to FBR cores from the design view points of thermal-hydraulic and core mechanical behaviors

Studies on the dissolution of mixed oxide spent fuel from FBR

International Nuclear Information System (INIS)

Nemoto, Shin-ichi; Shibata, Atsuhiro; Shioura, Takao; Okamoto, Fumitoshi; Tanaka, Yasumasa

1995-01-01

At the Chemical Processing Facility(CPF) in the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation(PNC), since 1982 Laboratory scale hot experiments have been carried out on the development of reprocessing technology for FBR mixed oxide fuel. The spent fuel pins which have been used in out experiments were irradiated in Experimental Fast Reactor 'Joyo' Phenix (France) and DFR(UK). Burn-up of the fuel pins were 4,400-100,000 MWd/t. This paper Summarizes a dissolution study that have been performed to define the Key parameters affecting dissolution rate such as concentration of nitric acid, burn-up, and temperature. And this paper also discusses about the character of releasing 85 Kr in chopping and dissolution process, and about the amount of insoluble residue. (author)

Practicalization strategic research of FBR cycle

International Nuclear Information System (INIS)

2000-01-01

Practicalization strategic research of FBR cycle consists of two phases such as phase I (FY 1999-2000) and phase II (to FY 2005). In every phase, research and development plants and results are checked and reviewed. The assessment indexes are five development objects such as safety, economical efficiency, resource effective utilization, environmental load decrease and nuclear non-proliferation and technical realization, too. Reactor core, FBR plant system and fuel cycle system are investigated. We selected the research subjects of cooling materials as sodium, heavy metals (lead and lead bismuth alloy), gas (carbon dioxide and helium) and water (boiling water, power water and supercritical pressure water) and fuel types as cladding tube fuel (oxide, nitride and metal) and coated fuel particle (oxide and nitride) for helium gas cooling reactor. In FY1999, the good reactor core and FBR plant system for every cooling materials are studied. Two reprocessing (a wet reprocessing using aqueous solution and a dry method) were selected. In FY 2000, we will investigate effects of throughput, plant concept and cost and evaluate achievement of development objects and then decide the development plan. (S.Y.)

A basic research on the transient behavior for a metallic fuel FBR

International Nuclear Information System (INIS)

Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

1999-03-01

A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

Studies in the dissolver off-gas system for a spent FBR fuel reprocessing plant

International Nuclear Information System (INIS)

Heinrich, E.; Huefner, R.; Weirich, F.

1982-01-01

Investigations of possible modifications of the process steps of a dissolver off-gas (DOG) system for a spent FBR fuel reprocessing plant are reported. The following operations are discussed: iodine removal from the fuel solution; behaviour of NOsub(x) and iodine in nitric acid off-gas scrubbers at different temperatures and nitric acid concentrations; iodine desorption from the scrub acid; selective absorption of noble gases in refrigerant-12; cold traps. The combination of suitable procedures to produce a total DOG system is described. (U.K.)

FBR/VHTR deployment scenarios in Japan

International Nuclear Information System (INIS)

Richards, Matt; Kunitomi, Kazuhiko

2008-01-01

Co-deployment of Fast Breeder Reactors (FBRs) and Very High Temperature Reactors (VHTRs) can be used as the nuclear technologies to meet a significant portion of Japan's future energy demands. The FBR provides the fissile fuel for energy security and sustainability, and can be used to provide a significant portion of the electricity demand. The VHTR can provide flexible energy outputs (electricity, hydrogen, and high-temperature heat) with high efficiency, can operate with a wide variety of fuel cycles, and can be sited at locations that have limited availability of cooling water. These features, combined with its passive safety and high degree of proliferation resistance, make the VHTR an ideal complement for co-deployment with the FBR in Japan and also a very low-risk technology of export to foreign countries. In addition to hydrogen production, the high-temperature thermal energy produced by the VHTR fleet can be used for a wide variety of process-heat applications, and the VHTR can play a key role for significantly reducing greenhouse-gas emissions. This paper describes assessments for deploying FBRs and VHTRs in Japan using a closed fuel cycle, with the FBRs supplying the fissile material to sustain the combined FBR/VHTR fleet. (author)

FBR and RBR particle bed space reactors

International Nuclear Information System (INIS)

Powell, J.R.; Botts, T.E.

1983-01-01

Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10 0 K), high coolant-outlet temperatures (1500 to 3000 0 K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H 2 -cooled mode. The RBR will operate only in the open-cycle H 2 -cooled mode

Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

International Nuclear Information System (INIS)

Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

2004-03-01

The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

FBR Plant Engineering Center annual report 2012

International Nuclear Information System (INIS)

2013-12-01

This annual report shows the last year's R and D activities of currently-reorganized FBR Plant Engineering Center, which was established on April 1, 2009. FBR Safety Technology Center was founded on April 1, 2013 by the consolidation of both the activities of 'former FBR Plant Engineering Center' and a portion of 'FBR Safety Evaluation Unit, Advanced Nuclear System Research and Development Directorate', especially concentrating on safety evaluations and analyses for severe accidents. As for FBR safety technology, it is necessary to continuously make an effort for compliance with new safety regulations in preparation for 'Monju' to restart, for safety enhancement evaluation and for safety technology upgrading. In this context, the new organization was founded in order to reinforce the safety evaluation capability, which will surely and steadily promote FBR safety-technology related activities. As a result, FBR Plant Engineering Center was abolished. This report summarizes the R and D activities at the former FBR Plant Engineering Center, aiming at contributing to the commercialization by using operation experiences and technology development results derived from the actual reactor 'Monju'. The activities are divided into five areas of operation-and-maintenance engineering, sodium engineering, reactor-core-and-fuel engineering, plant engineering, and safety engineering. This annual report is intended for a report of the activities of individual researcher in the center rather than that of the progress of the center as a whole. This will clarify the individual themes, progresses and problems of each researcher, which will, hopefully, facilitate communication with the outside researchers. (author)

Cost-benefit analysis on FBR cycle R and D for the world

International Nuclear Information System (INIS)

Kawasaki, Hirotsugu

2006-01-01

This analysis was estimated on the assumption that the nuclear power generation will be changed by FBR and both LWR and FBR indicate same nuclear power generation cost and the environmental load. The cost-benefit analysis results on FBR cycle R and D in the world showed that increase of power generation cost with increase of uranium fuel cost will be avoided and decrease of power generation cost by introducing FBR. The cost-benefit analysis results on FBR cycle R and D in Japan showed that about 9 billions yen will be obtained by the above two economic effects. Cost-benefit effects by introducing FBR, economic estimation method of cost-benefit effect, range and contents of cost-benefit effect on FBR R and D, preconditions of evaluation, and evaluation results are explained. (S.Y.)

Toward commercialization of FBR cycle (1). Promotion of R and D on technologies maintaining sustainable society

International Nuclear Information System (INIS)

Nagaoki, Yoshihiro; Nagura, Fuminori; Sakaguchi, Tomoyoshi; Kawasaki, Hirotsugu; Kikuchi, Shin

2008-01-01

The FBR cycle is a key technology maintaining a sustainable society through efficient utilization of limited uranium resources and conformance to global environmental protection. The domestic and overseas R and D of the FBR cycle entered on a new phase aiming at its commercialization and JAEA started the Fast Reactor Cycle Technology Development (FaCT) project. The FaCT project targeted at international standardization of the FBR fuel cycle and promoting the advanced R and D on the innovative technologies to increase cost-efficiency and reliability for the commercialization under international competition and cooperation. The combination of a sodium cooled FBR and advanced fuel cycle system with advanced aqueous reprocessing and simplified pelletizing fuel fabrication was selected a major concept. (T. Tanaka)

Method of determining the composition of fuels for FBR type reactors

International Nuclear Information System (INIS)

Tsutsumi, Kiyoshi.

1981-01-01

Purpose: To improve the core safety of FBR type reactors by determining the composition of fuels composed of oxide mixture of plutonium and uranium, using a relation between specific plutonium seed and plutonium enrichment degree. Method: Relation is determined between the ratio of a specific plutonium seed for constituting plutonium oxide, for example 239 U ratio and a plutonium enrichment degree required for setting the assembly power to a constant level. The ratio of 239 U is plutonium having a given isotopic ratio is also determined. The accuracy of the 239 U ratio can be improved by the correction using the density coefficient. Then, the plutonium enrichment degree is determined using the relation determined as above based on the thus determined 239 U ratio. The composition of the fuel using oxide mixture of plutonium and uranium is determined by utilizing the thus obtained plutonium enrichment degree. (Moriyama, K.)

Proposal of a nuclear cycle research and development plan in Tokai works. The roadmap from LWR cycle to FBR cycle

International Nuclear Information System (INIS)

Nakamura, Hirofumi; Abe, Tomoyuki; Kashimura, Takuo; Nagai, Toshihisa; Maeda, Seichiro; Yamaguchi, Toshiya; Kuroki, Ryoichiro

2003-07-01

The Generation-II Project Task Force Team has investigated a research and development plan of a future nuclear fuel cycle in Tokai works for about three months from December 19, 2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R and D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. Establishment and advancement of 'the tough nuclear fuel cycle'. Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste. And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology. (author)

FBR pellet fabrication - density and dimensional control

International Nuclear Information System (INIS)

Rasmussen, D.E.; Schaus, P.S.

1982-01-01

The fuel pellet fabricating experience described in this paper involved pellet processing tests using mixed oxide (PuO 2 -UO 2 ) powders to produce fast breeder reactor (FBR) fuel pellets. Objectives of the pellet processing tests were to establish processing parameters for sintered-to-size fuel pellets to be used in an irradiation test in the Fast Flux Test Facility and to establish baseline fabrication control information. 26 figures, 7 tables

Storage and disposal of high-level radioactive waste from advanced FBR fuel cycle

International Nuclear Information System (INIS)

Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

2011-01-01

Waste management of fast breeder reactor (FBR) fuel cycle with and without partitioning and transmutation (P and T) technology was investigated by focusing on thermal constraints due to heat deposition from waste in storage and disposal facilities including economics aspects of those facilities. Partitioning of minor actinides (MAs) and heat-generating fission products in high-level waste can enlarge the containment ratio of waste elements in the glass waste forms and shorten predisposal storage period. Though MAs can be transmuted in FBRs or dedicated transmuters, heat-generating fission products are difficult to be transmuted; they are partitioned and stored for a long time before disposal. The disposal concepts for heat-generating fission products and remainders such as rare-earth elements depend on storage period that ranges from several years to several hundreds of years. Short-term storage results in small size of storage facilities and large size of repositories, and vice versa for long-term storage. This trade-off relation was analyzed by estimating repository size as a function of storage period. The result shows that transmutation of MAs is essentially effective to reduce repository size regardless to storage period, and a combination of P and T can provide a smaller repository than the conventional one by two orders of magnitude. The cost analysis for waste management was also made based on rough assumptions on storage, transportation and repository excluding cost for introducing P and T that are still under evaluation. Cost of waste management for FBR without P and T is 0.25 Yen/kWh that is slightly smaller than that for LWR without P and T, 0.30 Yen/kWh. The introduction of MA transmutation to the FBR results in cost of 0.20 Yen/kWh, and full introduction of P and T provides the smallest cost of 0.08 Yen/kWh. (author)

Inner wall attack and its inhibition method for FBR fuel pin cladding at high burnup

International Nuclear Information System (INIS)

Xu Yongli; Long Bin; Li Jingang; Wan Jiaying

1998-01-01

The inner wall attack of the modified 316-Ti S.S. cladding tubes manufactured in China used FBR at 10at.% burnup was investigated by means of the out of pile simulation tests. The inner surface morphologies of the cladding tubes attached by fission products Cs, Te, I and Se at 700 deg. C under lower and high oxygen potentials were observed respectively, and the depth of attack was also measured. The burst strength, maximum circum expansion and the appearances of fracture were measured and observed respectively for the cladding tubes attacked by fission products. Based on the mechanism of FBR fuel cladding chemical interaction (FCCI), Cr, Zr and Nb were used as the oxygen absorbers respectively, in order to inhibit the inner wall attack of the cladding tubes. The corrosion morphologies and depth, the penetration depth of the fission products in the inner surface of the cladding tubes were detected. The inhibition effectiveness of the oxygen absorbers for the inner wall attack of the cladding tubes was evaluated. (author)

Role of analytical chemistry in the development of nuclear fuels

International Nuclear Information System (INIS)

Ramakumar, K.L.

2012-01-01

Analytical chemistry is indispensable and plays a pivotal role in the entire gamut of nuclear fuel cycle activities starting from ore refining, conversion, nuclear fuel fabrication, reactor operation, nuclear fuel reprocessing to waste management. As the fuel is the most critical component of the reactor where the fissions take place to produce power, extreme care should be taken to qualify the fuel. For example, in nuclear fuel fabrication, depending upon the reactor system, selection of nuclear fuel has to be made. The fuel for thermal reactors is normally uranium oxide either natural or slightly enriched. For research reactors it can be uranium metal or alloy. The fuel for FBR can be metal, alloy, oxide, carbide or nitride. India is planning an advanced heavy water reactor for utilization of vast resources of thorium in the country. Also research is going on to identify suitable metallic/alloy fuels for our future fast reactors and possible use in fast breeder test reactor. Other advanced fuel materials are also being investigated for thermal reactors for realizing increased performance levels. For example, advanced fuels made from UO 2 doped with Cr 2 O 3 and Al 2 O 3 are being suggested in LWR applications. These have shown to facilitate pellet densification during sintering and enlarge the pellet grain size. The chemistry of these materials has to be understood during the preparation to the stringent specification. A number of analytical parameters need to be determined as a part of chemical quality control of nuclear materials. Myriad of analytical techniques starting from the classical to sophisticated instrumentation techniques are available for this purpose. Insatiable urge of the analytical chemist enables to devise and adopt new superior methodologies in terms of reduction in the time of analysis, improvement in the measurement precision and accuracy, simplicity of the technique itself etc. Chemical quality control provides a means to ensure that the

Development of hydrogen production technology using FBR

International Nuclear Information System (INIS)

Ono, Kiyoshi; Otaki, Akira; Chikazawa, Yoshitaka; Nakagiri, Toshio; Sato, Hiroyuki; Sekine, Takashi; Ooka, Makoto

2004-06-01

This report describes the features of technology, the schedule and the organization for the research and development regarding the hydrogen production technology using FBR thermal energy. Now, the hydrogen production system is proposed as one of new business models for FBR deployment. This system is the production of hydrogen either thermal energy at approximately from 500degC to 550degC or electricity produced by a sodium cooled FBR. Hydrogen is expected to be one of the future clean secondary energies without carbon-dioxide emission. Meanwhile the global energy demand will increase, especially in Asian countries, and the energy supply by fossil fuels is not the best choice considering the green house effect and the stability of energy supply. The development of the hydrogen technology using FBR that satisfies 'sustainable energy development' and 'utilization of energies free from environmental pollution' will be one of the promising options. Based on the above mentioned recognition, we propose the direction of the development, the issues to be solved, the time schedule, the budget, and the organization for R and D of three hydrogen production technologies, the thermochemical hybrid process, the low temperature steam reforming process, and the high temperature steam electrolysis process in JNC. (author)

Status of feasibility study for various technical options of FBR systems

International Nuclear Information System (INIS)

Kani, Yoshio

2000-01-01

JNC (Japan Nuclear Cycle Development Institute) has started a new research project of feasibility studies (F/S) for a wide variety option of fast breeder reactor (FBR) and related fuel cycle in order to develop an economically competitive FBR cycle system fro commercialization. JNC and the electric untilities in Japan have established a new organization in JNC to perform the F/S since July 1, 1999. The organization has undertaken feasibility studies (F/S) in order to determine promising FBR cycle concepts and define necessary RandD tasks. The long-term targets of commercialized FBR cycle system are set as ensuring safety, economic competitiveness relative to future LWRs, efficient utilization of resources, reduction in environmental burden, and enhancement of nuclear non-proliferation. This paper describes the progress of design studies for a wide variety of technical options of FBR plants in the framework of the F/S. We make efforts towards considering all key issues so as not to fail to notice the best concept in a commercialized stage. In the study of technical options, the identified coolant types are sodium, heavy metal (lead and lead-bismuth), gas (carbon dioxide and helium ) and water (boiling water, pressurized water and supercritical water). The classified types of fuel are mixed oxide, nitride and metal. Design studies of small size modular plant concepts are also performed. We study many reactor concepts in combination with a coolant type and a fuel type, understand characteristics of each reactor concept based on our experience and an extensive survey of literature, and make a draft design of each reactor concept for rough estimation of construction costs. We also check how far the concept accomplishes each index (safety, economy, resource utilization, etc.) of design requirements, and will select several promising reactor concepts. (author)

Development of a FBR fuel bundle-duct interaction analysis code-BAMBOO. Analysis model and verification by Phenix high burn-up fuel subassemblies

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

2005-01-01

The bundle-duct interaction analysis code ''BAMBOO'' has been developed for the purpose of predicting deformation of a wire-wrapped fuel pin bundle of a fast breeder reactor (FBR). The BAMBOO code calculates helical bowing and oval-distortion of all the fuel pins in a fuel subassembly. We developed deformation models in order to precisely analyze the irradiation induced deformation by the code: a model to analyze fuel pin self-bowing induced by circumferential gradient of void swelling as well as thermal expansion, and a model to analyze dispersion of the orderly arrangement of a fuel pin bundle. We made deformation analyses of high burn-up fuel subassemblies in Phenix reactor and compared the calculated results with the post irradiation examination data of these subassemblies for the verification of these models. From the comparison we confirmed that the calculated values of the oval-distortion and bowing reasonably agreed with the PIE results if these models were used in the analysis of the code. (author)

Examination of fast reactor fuels, FBR analytical quality assurance standards and methods, and analytical methods development: irradiation tests. Progress report, April 1--June 30, 1976, and FY 1976

International Nuclear Information System (INIS)

Baker, R.D.

1976-08-01

Characterization of unirradiated and irradiated LMFBR fuels by analytical chemistry methods will continue, and additional methods will be modified and mechanized for hot cell application. Macro- and microexaminations will be made on fuel and cladding using the shielded electron microprobe, emission spectrograph, radiochemistry, gamma scanner, mass spectrometers, and other analytical facilities. New capabilities will be developed in gamma scanning, analyses to assess spatial distributions of fuel and fission products, mass spectrometric measurements of burnup and fission gas constituents and other chemical analyses. Microstructural analyses of unirradiated and irradiated materials will continue using optical and electron microscopy and autoradiographic and x-ray techniques. Analytical quality assurance standards tasks are designed to assure the quality of the chemical characterizations necessary to evaluate reactor components relative to specifications. Tasks include: (1) the preparation and distribution of calibration materials and quality control samples for use in quality assurance surveillance programs, (2) the development of and the guidance in the use of quality assurance programs for sampling and analysis, (3) the development of improved methods of analysis, and (4) the preparation of continuously updated analytical method manuals. Reliable analytical methods development for the measurement of burnup, oxygen-to-metal (O/M) ratio, and various gases in irradiated fuels is described

Fuel supply demand balances for future FBR commercialization: impacts on plutonium pricing and reactor design

International Nuclear Information System (INIS)

Braun, C.; Zebroski, E.L.

1985-01-01

Plutonium supply and demand balances for fast breeder reactor (FBR) commercialization post-2000 were computed to determine: (a) the maximum supportable number of FBRs that could be installed based on plutonium availability considerations and (b) the feasibility of a reasonable FBR capacity growth case assuming slow introduction post-2010 and rapid capacity growth post-2035. The purpose of the analysis was to determine the outer limitation on the maximum future FBR introduction, or the bounds of a possible plutonium-limited introduction rate, and to estimate the reasonableness of a more limited capacity growth case

The Swiss contribution to the FBR development

Energy Technology Data Exchange (ETDEWEB)

Hudina, M [Institut Federal de Recherches en Matiere de Reacteurs, Wuerenlingen (Switzerland); and others

1981-05-01

The program of fast breeder reactors development in Switzerland is considered from two points of view: energy self-sufficiency and optimization of fuel cycle. Research and development program covers: safety features of LMFBRs, development of mixed carbide fuel elements, study of steam generators transient behaviour, influence of various cooling concepts on thermal efficiency, techniques for detecting cover gas bubbles in the primary sodium circuit. This paper includes cost od the research and development activities as well as description of the future aims of the FBR projects.

The Swiss contribution to the FBR development

International Nuclear Information System (INIS)

Hudina, M.

1981-01-01

The program of fast breeder reactors development in Switzerland is considered from two points of view: energy self-sufficiency and optimization of fuel cycle. Research and development program covers: safety features of LMFBRs, development of mixed carbide fuel elements, study of steam generators transient behaviour, influence of various cooling concepts on thermal efficiency, techniques for detecting cover gas bubbles in the primary sodium circuit. This paper includes cost od the research and development activities as well as description of the future aims of the FBR projects

Preliminary study on flexible core design of super FBR with multi-axial fuel shuffling

International Nuclear Information System (INIS)

Sukarman; Yamaji, Akifumi; Someya, Takayuki; Noda, Shogo

2017-01-01

Preliminary study has been conducted on developing a new flexible core design concept for the Supercritical water-cooled Fast Breeder Reactor (Super FBR) with multi-axial fuel shuffling. The proposed new concept focuses on the characteristic large axial coolant density change in supercritical water cooled reactors (SCWRs) when the coolant inlet temperature is below the pseudocritical point and large coolant enthalpy rise is taken in the core for achieving high thermal efficiency. The aim of the concept is to attain both the high breeding performance and good thermal-hydraulic performance at the same time. That is, short Compound System Doubling Time (CSDT) for high breeding, large coolant enthalpy rise for high thermal efficiency, and large core power. The proposed core concept consists of horizontal layers of mixed oxide (MOX) fuels and depleted uranium (DU) blanket layers at different elevation levels. Furthermore, the upper core and the lower core are separated and independent fuel shuffling schemes in these two core regions are considered. The number of fuel batches and fuel shuffling scheme of the upper core were changed to investigate influence of multi-axial fuel shuffling on the core characteristics. The core characteristics are evaluated with-three-dimensional diffusion calculations, which are fully-coupled with thermal-hydraulics calculations based on single channel analysis model. The results indicate that the proposed multi-axial fuel shuffling scheme does have a large influence on CSDT. Further investigations are necessary to develop the core concept. (author)

FBR type reactor

International Nuclear Information System (INIS)

Hayase, Tamotsu.

1991-01-01

The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

The basic research on the CDA initiation phase for a metallic fuel FBR

International Nuclear Information System (INIS)

Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime

1998-03-01

A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

Study on reprocessing plant during transition period from LWR to FBR

International Nuclear Information System (INIS)

Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

2011-01-01

We have proposed a concept of a reprocessing plant suitable for the transition period from the light water reactors (LWRs) to the fast breeder reactors (FBRs) by making comparison of two plant concepts: (1) Independent Plant which processes LWR fuel and FBR fuel in separately constructed lines and (2) Modularized Plant which processes LWR fuel and FBR fuel in a same line. We made construction plans based on the reference power generation plan, and evaluated the Pu supply capability using the power generation plan as an indicator of plant operation flexibility. In general, a margin of processing capacity increases the Pu supply capability. The margin of the Modularized Plant necessary to obtain equivalent Pu supply capability is smaller than that of the Independent Plant. Also the margin of the Independent Plant results in decrease in the plant utilization factor. But the margin of the Modularized Plant results in little decrease in the plant utilization factor, because the Modularized Plant can address the types of reprocessing fuel to adjust to Pu demand and processing capacity. Therefore, the Modularized Plant has a greater potential for the reprocessing plants during transition period. (author)

Development of disassembly and pin chopping technology for FBR spent fuels

International Nuclear Information System (INIS)

Kobayashi, Tsuguyuki; Namba, Takashi; Kawabe, Yukinari; Washiya, Tadahiro

2008-01-01

Japan Atomic Power Company (JAPC) and Japan Atomic Energy Agency (JAEA) have been developing fuel disassembly and fuel pin chopping systems for a future Japanese commercial FBR. At first, the wrapper tube is cut by the slit-cut to pull it out, then the fuel pins are cut by the crop-cut at their end-plugs to separate them from the entrance nozzle. The pins are transferred to the magazine of the chopping machine. A series of tests were performed to develop this procedure. As the result of mechanical cutting tests, the CBN wheel was selected. The slit-cut tests were carried out to evaluated the cutting performance of the wheel. The wrapper tube is normally slit-cut in the circumferential direction. One CBN wheel could cut more than 5 fuel assemblies in this direction. The slit-cut in the axial direction is prepared as provision when the tube is difficult to put out. More work is needed to cut 5mm thick PNC-FMS plate in this direction without damaging the pins beneath it. As the result of the crop-cut tests of end-plugs made of ODS steel, the CBN wheel could cut the 61 pin bundle by two strokes. More work is needed to cut the 217 pin bundle. Fuel pin handling tests were performed to transfer them from the disassembly machine to the chopping machine. The Saucer tray was selected to receive the disassembled pins. All the pins were transferred and loaded into a magazine of the chopping machine. Fuel pin loading tests were conducted to optimize the magazine configuration to make the chopping length within 1.0±0.5 cm. In order to decrease the disturbance during chopping, the width of the magazine was adjusted to be 12 cm and installation of a height adjuster is favourable to control the free space above the pins. (author)

Parameters' influence estimation on Puf supply and demand in transitional period from LWR to FBR in Japan

International Nuclear Information System (INIS)

Kobayashi, Hiroaki; Ohta, Hirokazu; Inoue, Tadashi

2009-01-01

Plutonium fissile (Puf) amounts to balance supply and demand during transition period were evaluated with different parameters. Estimated total Puf demand in transitional period was sensitive to deployment speed of FBR. Because FBRs will be deployed as replacements of old LWRs for keeping total capacity, deployment history of existing LWRs should be taken into consideration. According to the estimation, LWR fuel burnup and utilized capacity are not big issue. Because certain amount of LWR spent fuel will remain in early phase of transitional period, there is enough time for preparing Puf supply. On the other hand, FBR fuel cycle time (SF cooling time + fuel fabrication time) have large impact on Puf supply. Fuel cycle technologies including transportation for applying to short cooling spent fuels should be developed. (author)

Study on hydrogen production using the fast breeder reactors (FBR)

International Nuclear Information System (INIS)

Kani, Yoshio

2003-01-01

As the fast breeder reactor (FBR) can effectively convert uranium-238 difficult to carry out nuclear fission at thermal neutron reactors to nuclear fissionable plutonium-239 to use it remarkable upgrading of application on uranium can be performed, to be expected for sustainable energy source. And, by reuse minor actinides of long half-life nuclides in reprocessed high level wasted solutions for fuels of nuclear reactors, reduction of radioactive poison based on high level radioactive wastes was enabled. As high temperature of about 800 centigrade was required on conventional hydrogen production, by new hydrogen production technique even at operation temperature of sodium-cooled FBR it can be enabled. Here were described for new hydrogen production methods applicable to FBR on palladium membrane hydrogen separation method carrying out natural gas/steam modification at reaction temperature of about 500 centigrade, low temperature thermo-chemical method expectable simultaneous simplification of production process, and electrolysis method expected on power load balancing. (G.K.)

Development of computer code SIMPSEX for simulation of FBR fuel reprocessing flowsheets: II. additional benchmarking results

International Nuclear Information System (INIS)

Shekhar Kumar; Koganti, S.B.

2003-07-01

Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)

Development of end plug welding method in the fabrication of FBR fuel pins

International Nuclear Information System (INIS)

Ohtani, Seiji; Sawayama, Takeo; Tateishi, Yoshinori

1977-01-01

As a part of the development of the automatic and remote controlled fabrication of FBR fuel pins, welding of fuel pin end plugs has been examined. Cladding tubes and end plugs used for this experiment are made of SUS 316, and they are the components of fuel pins for the prototype fast breeder reactor (Monju) or the second core of Joyo (Joyo MK-II). The welding tests of cladding tubes and four kinds of end plugs were carried out by means of two techniques; tungsten inert gas welding and laser welding. It can be said that no considerable difference was observed in weld penetration, occurrence rate of weld defects and breaking strength between the tight fit and the loose fit plugs. The face-to-face fit welding requires the least welding heat input, but involves much difficulty in the control of weld penetration and bead zone diameter. The good concentrative property and high energy density of laser beam make the face of weld hollow due to the vaporization of weld metal. However, this problem can be easily solved by changing the shape of end plugs. Good results in the other characteristics of the weld also were obtained by this laser welding. Further experiment is needed in connection with the compatibility of weld metal with sodium and neutron irradiation before final judgement is made on the laser welding technique. (Nakai, Y.)

Joint Feasibility study on the Instruction of a commercial FBR power plant to Korea between Korea and France

International Nuclear Information System (INIS)

Lee, Chang Kun; Cho, Mann; Juhn, Poong Eil; Nam, Jang Soo; Roh, Eun Rae; Yang, Chang Kook; Choi, Young Sang; Shin, Jae In; Chang, Young Nam; Kim, Hoa Gy; Lee, Hong Moon

1985-03-01

The objective of this joint feasibility study is to prepare for the implementation plan of the establishment of technological capability prior to the introduction of a commercial FBR power plant to Korea. The scope and contents of the second year project are as follows: Determined are the implementation activities in the two alternative FBR introduction scenarios i.e., a turnkey scenario and a gradual localization scenario with regard to the proportion of domestic participation. And FBR center is modeled as the construction of 4 units of 1,500 MWe LMFBR and related fuel cycle facilities such as reprocessing and fabrication plants at one site. In turnkey scenario, domestic implementation project is selected and the means to develop technical capability are classified into OJT/OJP, R and D, and technology transfer and the floating time schedule is made with reference to the first FBR operation year. In a gradual localization scenario Korean party's plan of means to develop technical capability is prepared. For the analysis of turnkey base introduction scenario project activities were extracted and analyzed referring to IAEA TR. 200 ''Manpower Development for Nuclear power Plant'', Guidebook (1980). And floating time schedule was prepared for the means to develop technical capability and project schedule. For the analysis of gradual localization scenario, project activities were derived under the assumption that entire facilities of FBR center should be localized. PWR's and FBR's fuel reprocessing facilities were incorporated in the reprocessing facility. (Author)

Evaluation of the commercial FBR introduction date

International Nuclear Information System (INIS)

White, M.K.; Merrill, E.T.

1981-09-01

This report examines one criterion for introducing a commercial FBR: economic competitiveness with a Light Water Reactor (LWR). For this analysis, the commercial FBR is assumed to be the fifth-of-a kind replicate which represents an economically mature plant. This FBR is deemed economically competitive when its life-cycle energy cost is less than or equal to that of an LWR. Results of this analysis are used in a comparative analysis of alternative FBR development stategies. The strategies evaluated in these studies assume both 1000- and 1457-MWe FBRs. Since the capital costs per kilowatt, and therefore the energy costs, for these two FBR sizes are different, they will become economically competitive at different times. The probability density function for the 1457-MW(e) FBR has an expected value date or weighted average date of 2030, compared with 2033 for the probability density function for the 1000-MW(e) FBR

Effects of applying three-dimensional seismic isolation system on the seismic design of FBR

International Nuclear Information System (INIS)

Hirata, Kazuta; Yabana, Shuichi; Kanazawa, Kenji; Matsuda, Akihiro

1997-01-01

In this study conceptional three-dimensional seismic isolation system for fast breeder reactor (FBR) is proposed. Effects of applying three-dimensional seismic isolation system on the seismic design for the FBR equipment are evaluated quantitatively. From the evaluation, it is concluded following effects are expected by applying the three-dimensional seismic isolation system to the FBR and the effects are evaluated quantitatively. (1) Reduction of membrane thickness of the reactor vessel (2) Suppression of uplift of fuels by reducing vertical seismic response of the core (3) Reduction of the supports for the piping system (4) Three-dimensional base isolation system for the whole reactor building is advantageous to the combined isolation system of horizontal base isolation for the reactor building and vertical isolation for the equipment. (author)

Development of a standard data base for FBR core nuclear design. 9. Analysis of FCA XVII-1 experiments

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto; Oigawa, Hiroyuki; Iijima, Susumu

1998-10-01

Pnc had developed the adjusted nuclear cross-section library in which the results of the Jupiter experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of Fbr cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In this report, the authors describe the evaluation of the C/E values and the sensitivity analysis for FCA XVII-1 assembly. FCA XVII-1 is a representative mock-up of a MOX fuel sodium cooling FBR core. The criticality, reaction rate ratio, sodium void reactivity worth and 238 U Doppler reactivity worth of FCA XVII-1 were analyzed. The results of C/E values calculated by the standard analytical method for JUPITER experiments are similar to those calculated by the method of JAERI, except for the sodium void reactivity. So, further investigation for sodium void reactivity is necessary. Furthermore, sensitivity analysis shows the characteristics of FCA XVII-1 in comparison with ZPPR-9. (author)

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Sakagami, M.

1984-01-01

FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

Scenario study on the FBR deployment

International Nuclear Information System (INIS)

Ono, Kiyoshi; Kofuji, Hirohide; Otaki, Akira; Yonezawa, Shigeaki; Shinoda, Yoshihiko; Hirao, Kazunori; Ikegami, Tetsuo

2000-12-01

This study on success scenarios for the Fast Breeder Reactor (FBR) deployment was performed taking account of future situation of fossil, renewable and nuclear energies in Japan as well as the world from the viewpoints of the following four items; economics, environment, energy security and restriction of natural uranium resources. In the economics scenario, if carbon tax is added to generating cost of LNG, coal and oil and the economics of FBR cycle is competitive with LWR cycle in the future, FBR cycle will be expected to introduce as the middle and base load power plant. In the environment scenario, there is also any possibility that FBR cycle which can burn and transmute minor actinide and fission product elements will be introduced in order to reduce the burden of deposit facility and the toxicity of high-level waste. In the uranium resources restriction scenario, FBR cycle needs to be deployed at the latest in the middle of 21st century from the viewpoint of the restriction of natural uranium resources. This study was carried out in a part of JNC's feasibility study on commercialized FBR cycle system. (author)

Method of operating FBR type reactors

International Nuclear Information System (INIS)

Arie, Kazuo.

1984-01-01

Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

Investigation of 144Pr measurements for determination of residual Pu in FBR leached hulls

International Nuclear Information System (INIS)

Yates, A.; Bremner, W.B.

1983-12-01

The measurement of plutonium in leached hulls arising from FBR fuel reprocessing is required for plant control, accountancy and safeguards. At DNPDE these measurements are carried out on the batch of hulls prior to bulking into 200 l drums for retrievable storage and ultimately further treatment for plutonium recovery. The experience to date has related to the use of neutron interrogation using sealed tube neutron generators as the irradiation source. The supply of replacement sealed tubes has become difficult. It was therefore decided to consider, amongst other techniques, the measurement of 144 Pr gamma emission as a possible alternative. The technique has been extensively used for thermal reactor fuels on a routine basis. There has also been a limited amount of work reported from Cap La Hague on applying the technique to FBR fuels as part of an experimental programme. This paper therefore describes the work done at DNPDE which evaluated the technique for use on a batch of hulls arising from a PFR fuel reprocessing campaign. (author)

Fast breeder fuel cycle

International Nuclear Information System (INIS)

1978-07-01

This contribution is prepared for the answer to the questionnaire of working group 5, subgroup B. B.1. is the short review of the fast breeder fuel cycles based on the reference large commercial Japanese LMFBR. The LMFBRs are devided into two types. FBR-A is the reactor to be used before 2000, and its burnup and breeding ratio are relatively low. The reference fuel cycle requirement is calculated based on the FBR-A. FBR-B is the one to be used after 2000, and its burnup and breeding ratio are relatively high. B.2. is basic FBR fuel reprocessing scheme emphasizing the differences with LWR reprocessing. This scheme is based on the conceptual design and research and development work on the small scale LMFBR reprocessing facility of Japan. The facility adopts a conventional PUREX process except head end portions. The report also describes the effects of technical modifications of conventional reprocessing flow sheets, and the problems to be solved before the adoption of these alternatives

Base technology development of new materials for FBR performance innovations

International Nuclear Information System (INIS)

Kano, Shigeki; Koyama, Masahiro; Nomura, Shigeo; Morikawa, Satoru; Ueno, Fumiyoshi

1989-01-01

This paper describes the base technology development of new materials for FBR performance innovations at the Power Reactor and Nuclear Fuel Development Corporation. The contents are as follows: (1) development of sodium and radiation resistant new materials, (2) development of high performance shielding material, (3) development of high performance control material, (4) development of new functional materials for reactor instrumentation. (author)

Study on design method for seismically isolated FBR plants

International Nuclear Information System (INIS)

Hirata, Kazuta; Yabana, Shuichi; Ohtori, Yasuki; Ishida, Katsuhiko; Sawada, Yoshihiro; Shiojiri; Hiroo; Mazda, Taiji

1998-01-01

CRIEPI conducted 'Demonstration test on FBR seismic isolation system' from 1987 to 1996 under contract with Ministry of International Trade and Industry, Japan. In the demonstration test, base isolation technologies are prepared and demonstrated to apply to FBR and the design guidelines are proposed. In this report overall contents of the design guidelines entitled Design guidelines for seismically base isolated FBR plants' are included. The design guidelines, as a rule, are limited to apply to FBR plants where entire reactor building is isolated in the horizontal direction using laminated rubber bearings as isolators. The design guidelines and its concepts, however, will be useful for the development of similar guidelines for other isolation systems using different type of isolation methods and other nuclear facilities. The design guidelines consist of three parts and appendices. The first part is 'Policy for Safety Design of Base Isolated FBR Plants' specifying the principles and the requirements in the planning and the design for the safety of base isolated FBR plants. The second part is Policy for Seismic Design of Base Isolated FBR' describing the principles and the requirements in the seismic design and the evaluation of safety for base isolated FBR plants. The third part is 'Design Methods for Seismic Isolated FBR Plants' detailing the methods, procedures and parameters to be used in the design and the evaluation of safety fro base isolated FBR plants. In appendices examples of design procedures for base isolated reactor building and laminated rubber bearings as well as various test data on laminated rubber bearings, etc. are shown. (author)

Nondestructive analysis of irradiated fuels

International Nuclear Information System (INIS)

Dudey, N.D.; Frick, D.C.

1977-01-01

The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

Method of controlling power distribution in FBR type reactors

International Nuclear Information System (INIS)

Sawada, Shusaku; Kaneto, Kunikazu.

1982-01-01

Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

Engineering scale tests of mechanical disassembly and short stroke shearing systems for FBR fuel assembly

International Nuclear Information System (INIS)

Higuchi, Hidetoshi; Kitagaki, Toru; Koizumi, Kenji; Hirano, Hiroyasu; Takeuchi, Masayuki; Washiya, Tadahiro; Kawabe, Yukinari; Kobayashi, Tsuguyuki

2011-01-01

Japan Atomic Energy Agency (JAEA) and The Japan Atomic Power Company (JAPC) have been developing an advanced head-end process based on mechanical disassembly and short stroke shearing systems as a part of Fast Reactor Cycle Technology Development (FaCT). Fuel pins for a fast reactor are installed within a hexagonal shaped wrapper tube made of stainless steel. In order to reprocess the fast reactor fuel pins, they must be removed from the wrapper tube and transported to the shearing system without failure. In addition, the advanced aqueous reprocessing process, called 'NEXT' (New Extraction System for TRU Recovery) process requires a solution of the spent fuel with relatively high concentration (500g/L). JAEA and JAPC have developed the mechanical disassembly and the short stroke shearing technology which is expected to make fragmented fuel to satisfy these requirements. This paper reports the results of engineering scale tests on the mechanical disassembly and short stroke shearing systems. These tests were carried out with simulated FBR fuel assembly and removed pins. The mechanical cutting method has been developed to avoid fuel pin failure during disassembly operation. The cutting process is divided into two modes, so called 'slit-cut' for cutting the wrapper tube and 'crop-cut' for the end plug region of the fuel pin bundle. In the slit-cut mode, the depth of cutting was automatically controlled based on the calculated wastage of the cutting tool and deformation of the wrapper tube which had been measured before the cutting. This procedure was confirmed to minimize the fuel pin failure which was hard to prevent in the case of laser cutting. The cutting speed was also controlled automatically by the electric current of the cutting motor to lower the load of the cutting tool. The removed fuel pins were transported to the shearing machine, whose fuel shearing magazine width was set to be narrow to realize the suitable configuration for the short stroke shearing

General purpose nonlinear analysis program FINAS for elevated temperature design of FBR components

International Nuclear Information System (INIS)

Iwata, K.; Atsumo, H.; Kano, T.; Takeda, H.

1982-01-01

This paper presents currently available capabilities of a general purpose finite element nonlinear analysis program FINAS (FBR Inelastic Structural Analysis System) which has been developed at Power Reactor and Nuclear Fuel Development Corporation (PNC) since 1976 to support structural design of fast breeder reactor (FBR) components in Japan. This program is capable of treating inelastic responses of arbitrary complex structures subjected to static and dynamic load histories. Various types of finite element covering rods, beams, pipes, axisymmetric, two and three dimensional solids, plates and shells, are implemented in the program. The thermal elastic-plastic creep analysis is possible for each element type, with primary emphasis on the application to FBR components subjected to sustained or cyclic loads at elevated temperature. The program permits large deformation, buckling, fracture mechanics, and dynamic analyses for some of the element types and provides a number of options for automatic mesh generation and computer graphics. Some examples including elevated temperature effects are shown to demonstrate the accuracy and the efficiency of the program

Evaluation of bundle duct interaction by out-of-pile compression test of FBR fuel pin bundles

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Kosuke; Yamamoto, Yuji; Nagamine, Tsuyoshi; Maeda, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

2001-06-01

Bundle duct interaction (BDI) caused by expansion of fuel pin bundle is a main factor to limit the fuel lifetime. Therefore, it is important for the design of fast reactor fuel assembly to understand the fuel pin deformation behavior under BDI condition. In order to understand the fuel pin deformation behavior under BDI condition, out-of-pile compression tests were conducted for FBR fuel pin bundle by use of X-ray CT equipment. In these compression tests, two kinds of fuel pin bundles were conducted. One was the fuel pin bundle with the short wire-pitch and the other was the fuel pin bundle with the short wire-pitch and large diameter claddings. The general discussions were also performed based on the results of out-of-pile compression tests obtained by use of X-ray CT equipment in the previous work. Following results were obtained. 1) The occurrence of the pin-to-duct contact depends on the wire-pitch. In the fuel pin bundle with large wire-pitch, the pin-to-duct contact occurred at the early stage of BDI. The reason of this result is due to the low bowing rigidity of the fuel pins with long wire-pitch. 2) The value of the ovalation stiffness strongly depends on the geometry of cladding (diameter, thickness) and especially on wire-pitch. This result in this work revealed that the occurrence of the pin-to-duct contact depends on the value of the ovalation stiffness. 3) The occurrence of wire dispersion and dispersive displacement of pins depends on the wire-pitch strongly. In the fuel pin bundle with the long wire-pitch, the occurrence of the above-mentioned suppression mechanism to BDI is remarkable. 4) The suppression mechanism to BDI of the fuel pin bundle with the long wire-pitch is elastic oval deformation of cladding, wire dispersion and dispersive displacement of pins. On the other hand, the elastic and plastic oval deformation of cladding is the major suppression mechanism to BDI in the fuel pin bundle with the short wire-pitch. 5) The appearance of

Summary: analysis of alternative FBR development strategies

International Nuclear Information System (INIS)

Burnham, J.B.

1981-12-01

This report summarizes the comparative evaluation of alternative strategies for the development of the commercial fast breeder reactor (FBR) in the United States. For planning purposes, a range of possible FBR development paths called strategies were selected for evaluation. These strategies, designed to be technically and economically feasible, were expressed in terms of the timing and nature of facilities/research and development programs required to reach full power operation of the first commercial FBR. Four of the seven strategies resulted in a large (1457 MWe) FBR as an end point, the other three in a 1000-MWe plant. Probability distributions were calculated for total strategy costs and time to completion. For the seven strategies analyzed, the costs (discounted 1980 dollars) ranged from $1.8 billion to $4.9 billion; the completion times ranged from 24 to 55 years

Present condition of survey research on actualization strategy of fast breeder reactor (FBR) cycling. Design research on fuel production system

International Nuclear Information System (INIS)

Tanaka, Kenya

2001-01-01

The fuel production system design investigation was performed for construction of fuel production process concept and plant image searching for the targets such as economics, environmental loading reduction, and so on required for practical use of FBR fuel recycling at a premise of safety security. By expectation of economics as a fuel cycling system, enhancement of nuclear proliferation resistance, and so on, it becomes more important to investigate on a fuel cycling system suitable for raw materials with low decontamination and high radiation intensity. In addition, it is also necessary to carry out investigation on fuel production system concept accompanies with MA recycling system for reduction of environmental loading. Therefore, investigation objects on the system were laid their fundamental processes on denitrification conversion/pelletizing process and gelation/vibration filling process for raw material solution from advancing wet reprocessing and on vibration filling process for oxide granules obtained from dry reprocessing system and casting method for metal fuels. As a result, for the pollution removal fuel production system suitable for either of wet/dry reprocessing, a mass-production scale production plant image was elucidated at a premise of production yield, realizability of remote automation system, and so on. On candidate concepts of every fuel production system, no fatal defect was found on results of outline evaluation on features of system such as production facility scale and so on before present stage. (G.K.)

Advances in nuclear fuel technology. 3. Development of advanced nuclear fuel recycle systems

International Nuclear Information System (INIS)

Arie, Kazuo; Abe, Tomoyuki; Arai, Yasuo

2002-01-01

Fast breeder reactor (FBR) cycle technology has a technical characteristics flexibly easy to apply to diverse fuel compositions such as plutonium, minor actinides, and so on and fuel configurations. By using this characteristics, various feasibilities on effective application of uranium resources based on breeding of uranium of plutonium for original mission of FBR, contribution to radioactive wastes problems based on amounts reduction of transuranium elements (TRU) in high level radioactive wastes, upgrading of nuclear diffusion resistance, extremely upgrading of economical efficiency, and so on. In this paper, were introduced from these viewpoints, on practice strategy survey study on FBR cycle performed by cooperation of the Japan Nuclear Cycle Development Institute (JNC) with electric business companies and so on, and on technical development on advanced nuclear fuel recycle systems carried out at the Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute, and so on. Here were explained under a vision on new type of fuels such as nitride fuels, metal fuels, and so on as well as oxide fuels, a new recycle system making possible to use actinides except uranium and plutonium, an 'advanced nuclear fuel cycle technology', containing improvement of conventional wet Purex method reprocessing technology, fuel manufacturing technology, and so on. (G.K.)

Study on commercial FBR concepts by combining innovative technologies

International Nuclear Information System (INIS)

Miura, M.; Inagaki, T.; Kuroha, M.; Hida, T.

1992-01-01

A study was conducted on future prospects of FBR commercialization. Targets of further improving safety and economy were set to make commercial power plants that would be superior to future LWRs. Promising innovative technologies studied domestically and overseas were extracted by evaluating prospects for commercialization, effect, and plant applicability. Several commercial plants were conceptualized by introducing such technology to large-scale and oxide-fuel reactors. Estimates of construction cost, etc., proved that the targets could be achieved. A concept of long-term technological development was synthesized. (author)

Analysis of nuclear fuel utilization strategies

International Nuclear Information System (INIS)

Souza, J.A.M. de; Rubini, L.A.; Silva, R.P.

1981-01-01

Having in mind a rational utilization of the national uranium reserves, in the alternatives considered here in both the recycling of irradiated fuel and the introduction of fast breeder reactors (FBR's) in the electric energy generation system are analyzed. Present thermal reactors (PWR's) would be, thus, gradually replaced, and the plutonium produced by them used as fuel for the next generation of FBR's. (Author) [pt

Comparison of seismic isolation concepts for FBR

International Nuclear Information System (INIS)

Shiojiri, H.; Mazda, T.; Kasai, H.; Kanda, J.N.; Kubo, T.; Madokoro, M.; Shimomura, T.; Nojima, O.

1989-01-01

This paper seeks to verify the reliability and effectiveness of seismic isolation for FBR. Some results of the preliminary study of the program are described. Seismic isolation concepts and corresponding seismic isolation devices were selected. Three kinds of seismically-isolated FBR plant concepts were developed by applying promising seismic isolation concepts to the non-isolated FBR plant, and by developing plant component layout plans and building structural designs. Each plant was subjected to seismic response analysis and reduction in the amount of material of components and buildings were estimated for each seismic isolation concepts. Research and development items were evaluated

Creep fatigue design of FBR components

International Nuclear Information System (INIS)

Bhoje, S.B.; Chellapandi, P.

1997-01-01

This paper deals with the characteristic features of Fast Breeder Reactor (FBR) with reference to creep fatigue, current creep fatigue design approach in compliance with RCCMR (1987) design code, material data, effects of weldments and neutron irradiation, material constitutive models employed, structural analysis and further R and D required for achieving maturity in creep fatigue design of FBR components. For the analysis reported in this paper, material constitutive models developed based on ORNIb (Oak Ridge National Laboratory) and Chaboche viscoplastic theories are employed to demonstrate the potential of FBR components for higher plant temperatures and/or longer life. The results are presented for the studies carried out towards life prediction of Prototype Fast Breeder Reactor (PFBR) components. (author). 24 refs, 8 figs, 5 tabs

Low-decontamination approach to a proliferation-resistant fuel cycle

International Nuclear Information System (INIS)

Asquith, J.G.; Grantham, L.F.

1978-01-01

To prevent the diversion of nuclear material from power production to weapon production either by a nation or by clandestine groups within a nation, the nuclear fuel cycle must be proliferation-resistant and safeguarded. Potentially proliferation-resistant and safeguarded fuel cycles based on low-decontamination pyroreprocessing have been developed for the light water reactor (LWR), fast breeder reactor (FBR), and FBR-LWR combination. The major penalty for recycling fission products to the LWR is that fuel enrichment must be somewhat greater to overcome parasitic fission product absorption of neutrons. In the FBR, the major penalty is a slight reduction in breeding ratio due to the displacement of fertile material by fission products. Preliminary cost analysis indicates that these fuel cycles are economically competitive with fuel cycles using conventional reprocessing or those using virgin uranium if spent fuel storage costs are considered

Pu Denaturing by Transmutation of MA in FBR Multi-cycle

Energy Technology Data Exchange (ETDEWEB)

Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi [Tokyo Institute of Technology, 2-12-1-N1-1 Ookayama, Meguro-ku, Tokyo, 1528550 (Japan)

2009-06-15

Pu accumulation and its recycling is important in the term of energy resources, however one of the most sensitive issues is non-proliferation in the future fuel cycle based on fast breeder reactor (FBR). The present paper utilizes Protected Pu Production (P{sup 3}) concept for the production of {sup 238}Pu and {sup 242}Pu by Minor Actinides (MA) transmutation to enhance the proliferation resistance of Pu in the fuel. Increase in the {sup 238}Pu and {sup 242}Pu isotopic fraction creates a high rate of internal heat generation by alpha decay (DH) and/or a high neutron source of spontaneous fission (SFN) in Pu that would be encountered during manufacturing and maintaining of nuclear explosive device. The feasibility of denaturing of Pu by MA transmutation in medium size FBR has been studied from the viewpoint of even-mass number Pu accumulation during multi-cycle of Pu and MA. The proliferation resistance property of Pu is also evaluated based on the specific decay heat and spontaneous fission neutron, compared with the reference criteria. In present paper, the P{sup 3} technology based on multi-recycled Pu and MA is compared with the conventional technology based on multi-recycled Pu only. The detail of mass balance behavior is, however, beyond the scope of the present paper. (authors)

Transition period fuel cycle from current to next generation reactors for Japan

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2007-01-01

Long-term energy security and global warming prevention can be achieved by a sustainable electricity supply with next generation fast breeder reactors (FBRs). Current light water reactors (LWRs) will be replaced by FBRs and FBR cycle will be established in the future considering the limited amount of uranium (U) resource. The introduction of FBRs requires plutonium (Pu) recovered from LWR spent fuel. The authors propose advanced system named Flexible Fuel Cycle Initiative (FFCI)' which can supply enough Pu and hold no surplus Pu, can respond flexibly the future technical and social uncertainties, and can achieve an economical FBR cycle. FFCI can simplify the 2nd LWR reprocessing facility for Japan (after Rokkasho Reprocessing Plant) which only carries out U removal from LWR spent fuel. Residual 'Recycle Material' is, according to FBRs introduction status, immediately treated in the FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of Recycle Material as a buffer material between LWR and FBR cycles. (author)

Overview of MOX fuel fabrication achievements

International Nuclear Information System (INIS)

Bairiot, H.; Vliet, J. van; Chiarelli, G.; Edwards, J.; Nagai, S.H.; Reshetnikov, F.

2000-01-01

Such overview having been adequately covered in an OECD/NEA publication providing the situation as of end 1994, this paper is mainly devoted to an update as of end 1998. The Belgian plant, Belgonucleaire/Dessel, is now dedicated exclusively to the fabrication of MOX fuel and has operated consistently around its nameplate capacity (35tHM/a) through the 1990s involving a large variety of PWR and BWR fuels. The two French plants have also achieved routine operation during the 1990s. CFCa, historically the largest FBR MOX fuel manufacturer, is utilizing the genuine COCA process for that type of fuel and the MIMAS process for LWR fuel: a nominal capacity (40 tHM/a) has been gradually approached. MELOX has operated at 100 tHM/a, as defined in the operating licence granted originally. The British plant, MDF/Sellafield with 8tHM/a nameplate capacity is devoted to fuel and has manufactured several small fabrication campaigns. In Japan, JNC operates three facilities located at Tokai: PFDF, devoted to basic research and fabrication of test fuels, PFFF/ATR line, for the fabrication of Fugen fuel and of corresponding fuel for the critical facility DCA, and PFPF for the fabrication of FBR fuel. In Russia, fabrication techniques have been developed to fuel four BN-800 FBRs contemplated to be constructed and be fuelled with the civilian Pu stockpile. Two demonstration facilities Paket (Mayak) and RIAR (Dimitrovgrad) fabricated respectively pellet and vipac type FBR MOX fuel for BR-5, BOR-60, BN-350 and BN-600. The paper includes a brief description of each of the fabrication routes mentioned, as well as the production of respectively LWR and FBR MOX fuel in each fabrication facility, since the start-up of the plant, since 1 January 1993 and since 1 January 1998 up to 31 December 1998. (author)

Development of a FBR fuel pin bundle deformation analysis code 'BAMBOO' . Development of a dispersion model and its validation

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ukai, Shigeharu; Asaga, Takeo

2002-03-01

Bundle Duct Interaction (BDI) is one of the life limiting factors of a FBR fuel subassembly. Under the BDI condition, the fuel pin dispersion would occur mainly by the deviation of the wire position due to the irradiation. In this study the effect of the dispersion on the bundle deformation was evaluated by using the BAMBOO code and following results were obtained. (1) A new contact analysis model was introduced in BAMBOO code. This model considers the contact condition at the axial position other than the nodal point of the beam element that composes the fuel pin. This improvement made it possible in the bundle deformation analysis to cause fuel pin dispersion due to the deviations of the wire position. (2) This model was validated with the results of the out-of-pile compression test with the wire deviation. The calculated pin-to-duct and pin-to-pin clearances with the dispersion model almost agreed with the test results. Therefore it was confirmed that the BAMBOO code reasonably predicts the bundle deformation with the dispersion. (3) In the dispersion bundle the pin-to-pin clearances widely scattered. And the minimum pin-to-duct clearance increased or decreased depending on the dispersion condition compared to the no-dispersion bundle. This result suggests the possibility that the considerable dispersion would affect the thermal integrity of the bundle. (author)

Nuclear fuels and development of nuclear fuel elements

International Nuclear Information System (INIS)

Sundaram, C.V.; Mannan, S.L.

1989-01-01

Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

Technical study report on fuel fabrication system

International Nuclear Information System (INIS)

Kono, Shusaku; Tanaka, Kenya; Ono, Kiyoshi; Iwasa, Katsuyoshi; Hoshino, Yasushi; Shinkai, Yasuo

2000-07-01

The feasibility study of FBR and related fuel cycle is performed for developing the FBR recycle system which ensures safety, economic competitiveness, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation under consistency of FBR reactor and fuel cycle systems. In this study, a conceptual design study and system characteristics evaluation are conducted for fuel fabrication systems of pellet process, vibropack process for oxide and nitride fuel and casting process for metal fuel. Technical issues in each process are also extracted. In 1999 fiscal year, a conceptual design study were conducted for the fuel fabrication plants adopting (1) the short pellet process which simplifies the conventional MOX pellet fabrication processes, (2) vibropack processes of aqueous gelation process, improved RIAR process, improved ANL process and fluoride volatility process, (3) casting processes of injection process, centrifuging process. As a result, attainable perspective was obtained for each fuel fabrication system through the evaluation of apparatuses, layout and facility volume, etc. In each fuel fabrication system, technical issues for practical use were made clear. Hereafter, more detailed study will be performed for each system, and research programs for phase II study will be planned. (author)

Experience with, and programme of, FBR and HWR development in Japan

International Nuclear Information System (INIS)

Iida, M.; Sawai, S.; Nomoto, S.

1983-01-01

Nuclear power generation in Japan is moving forward on the long-term development programme of nuclear power from the LWR to the FBR, essentially in the same way as in other advanced nuclear countries. In this development programme the unique HWR is also included; it can use plutonium produced in LWRs together with depleted uranium before the introduction of commercial FBRs. This report describes the status of the FBR and HWR development project being carried out by the Power Reactor and Nuclear Fuel Development Corporation (PNC) based upon the Long-Term Programme on Research, Development and Utilization of Nuclear Energy in Japan. Operational experience and technical results are shown for the experimental fast reactor JOYO (100 MW(th)), which reached initial criticality in 1977. The status of the 280 MW(e) prototype reactor MONJU, under construction as of 1982, is described. The conceptual design of the subsequent 1000 MW(e) demonstration plant is outlined, as is additional future planning. Research and development results, mainly carried out at Oarai Engineering Center of PNC, are shown. The 165 MW(e) prototype FUGEN is a heavy-water-moderated, boiling-light-water-cooled, pressure-tube-type reactor which uses plutonium mixed-oxide fuel. This report describes the relationship of the fuel cycle to the HWR in Japan and also discusses the operational experience of the prototype FUGEN, which has operated since 1979. Also described is the design of the 600 MW(e) demonstration plant and the programme of related research and development. (author)

Development of probabilistic fast reactor fuel design method

International Nuclear Information System (INIS)

Ozawa, Takayuki

1997-01-01

Under the current method of evaluating fuel robustness in FBR fuel rod design, a variety of uncertain quantities including fuel production tolerance and power density are estimated conservatively. In the future, in order to proceed with improvements in the FBR core's performance and optimize the fuel's specifications, a rationalization of fuel design tolerance is required. Among the measures aimed at realizing this rationalization, the introduction of a probabilistic fast reactor fuel design method is currently under consideration. I have developed a probabilistic fast reactor fuel design code named BORNFREE, in order to make use of this method in FBR fuel design. At the same time, I have carried out a trial calculation of the cladding stress using this code and made a study and an evaluation of the possibility of employing tolerance rationalization in fuel rod design. In this paper, I provide an outline description of BORNFREE and report the results of the above study and evaluation. After performing cladding stress trial calculations using the probabilistic method, I was able to confirm that this method promises more rational design evaluation results than the conventional deterministic method. (author)

Fuel exchanger in FBR type reactor

International Nuclear Information System (INIS)

Shinden, Kazuhiko; Tanaka, Osamu.

1990-01-01

The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

Success tree analysis on the technologies development for FBR commercialization

International Nuclear Information System (INIS)

An, Shigehiro; Taniyama, Hiroshi; Nagai, Hiroshi.

1991-01-01

In order to obtain a secure energy supply in future, it is important to establish a system for plutonium utilization via the FBR which is superior to the uranium utilization system with respect to both safety and good economics. In spite of this obvious need, the commercialization of the FBR is facing delays. Although several factors, for example, improvement of LWR technologies, stable supply of low cost uranium, opposition to nuclear power, etc. are contributors, the primary reason for the delay is the unfavorable economics of the FBR itself. In this paper the key technologies leading to reduced FBR costs are identified and their development strategies are discussed. (author)

Laser-based analytical monitoring in nuclear-fuel processing plants

International Nuclear Information System (INIS)

Hohimer, J.P.

1978-09-01

The use of laser-based analytical methods in nuclear-fuel processing plants is considered. The species and locations for accountability, process control, and effluent control measurements in the Coprocessing, Thorex, and reference Purex fuel processing operations are identified and the conventional analytical methods used for these measurements are summarized. The laser analytical methods based upon Raman, absorption, fluorescence, and nonlinear spectroscopy are reviewed and evaluated for their use in fuel processing plants. After a comparison of the capabilities of the laser-based and conventional analytical methods, the promising areas of application of the laser-based methods in fuel processing plants are identified

Advanced fuel for fast breeder reactors: Fabrication and properties and their optimization

International Nuclear Information System (INIS)

1988-06-01

The present design for FBR fuel rods includes usually MOX fuel pellets cladded into stainless steel tubes, together with UO 2 axial blanket and stainless steel hexagonal wrappers. Mixed carbide, nitride and metallic fuels have been tested as alternative fuels in test reactors. Among others, the objectives to develop these alternative fuels are to gain a high breeding ratio, short doubling time and high linear ratings. Fuel rod and assembly designers are now concentrating on finding the combination of optimized fuel, cladding and wrapper materials which could result in improvement of fuel operational reliability under high burnups and load-follow mode of operation. The purpose of the meeting was to review the experience of advanced FBR fuel fabrication technology, its properties before, under and after irradiation, peculiarities of the back-end of the nuclear fuel cycle, and to outline future trends. As a result of the panel discussion, the recommendations on future Agency activities in the area of advanced FBR fuels were developed. A separate abstract was prepared for each of the 10 presentations of this meeting. Refs, figs and tabs

Assessment of the thorium and uranium fuel cycle in the fast breeder and the high temperature reactor

International Nuclear Information System (INIS)

Schikorr, W.M.

1977-01-01

This report assesses the fissile fuel economy of the uranium and thorium cycle in the advanced reactors currently under development, the fast breeder reactor (FBR) and the high temperature reactor (HTR). It is shown by means of detailed burnup calculations that replacing UO 2 with ThO 2 or Th-metal as the radial blanket breeding material will not have any significant imapct on the breeding and burnup properties of the FBR. A global, analytical investigation is performed to study the fissile fuel economy of the many fissile fuel cycles possible in the HTR. Here it is demonstrated that the optimum conversion ratio of CR 3 O 8 ) demands are evaluated for a country such as the FRG under the assumptions of different future reactor strategy scenarios. Here it is demonstrated that the employement of both HTRs and FBRs can lead to a practically resource independent energy supply system within the next 40 to 60 years. However only through the large scale employement of the fast breeder can the future nuclear resource requirements be assured. (orig.) [de

A strategy analysis of the fast breeder reactor introduction and nuclear fuel cycle systems deployment

International Nuclear Information System (INIS)

Wajima, Tsunetaka; Kawashima, Katsuyuki; Yamashita, Takashi

1996-01-01

A study is made on a strategy analysis of the long term nuclear fuel cycle systems deployment in accordance with the nuclear power growth projection and fast breeder reactor (FBR) introduction. In the analysis, the reprocessed plutonium (Pu) is charged into the reactor in such a way that the reprocessed Pu is not stored outside the reactor, i.e., there is no excess Pu outside the reactor. The analysis characterized the fuel cycle systems, and showed the usefulness of the present method to determine future directions for the FBR introduction and nuclear fuel cycle systems deployment. Concerning an intermediate-term strategy, the time of introduction and required capacities of a second commercial LWR reprocessing plant, Pu-thermal, and the first FBR reprocessing plant deployment are evaluated. A long term strategy analysis shows that the two or three large plants are run in parallel for each fuel cycle facility and that FBR related facilities deal with a markedly large amount of Pu. It is concluded that the early stage introduction of FBRs of significant capacities seems necessary to materialize a consistent total FBR/fuel cycle system where Pu balance becomes feasible through its flexible operation of, for instance, adjusting breeding ratio, in order to keep the transparency of the Pu utilization. (author)

The uranium-plutonium breeder reactor fuel cycle

International Nuclear Information System (INIS)

Salmon, A.; Allardice, R.H.

1979-01-01

All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

Main lessons for FBR safety study from the CABRI experiments

International Nuclear Information System (INIS)

Sato, Ikken

2006-01-01

CABRI project has been carried out FBR safety study with international cooperation of five nations since 1978. The project consists of four periods such as CABRI-1, CABRI-2, CABRI-FAST and CABRI-RAFT. The objects and the main information for hypothetical core decay accident and safety study of fuel are described. The behavior of core decay accident was studied in the first period (CABRI-1), the safety study of fuel was investigated after the second period (CABRI-2). Change of phenomena at the initial process of core decay accident, comparison between analysis and data of fuel diffusion behavior of neutron horoscope by CABRI-2 experiment, representative in-core experiment under slow TOP conditions, the damaged and undamaged pin under slow TOP conditions, the exterior of CABRI and TREAT and the upper part of CABRI and TREAT reactor are shown. CABRI experiments changed to LWR safety study and a part of TREAT is stated. (S.Y.)

Development of 3-D FBR heterogeneous core calculation method based on characteristics method

International Nuclear Information System (INIS)

Takeda, Toshikazu; Maruyama, Manabu; Hamada, Yuzuru; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro

2002-01-01

A new 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed by combining the characteristics method and the nodal transport method. In the axial direction the nodal transport method is applied, and the characteristics method is applied to take into account the radial heterogeneity of fuel assemblies. The numerical calculations have been performed to verify 2-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction rates in a small region

Experimental study associated to irradiation of FBR structural material, (4)

International Nuclear Information System (INIS)

1976-01-01

The study presents one of the bases to evaluate the results of the post-irradiation tests to conduct the thermal control tests related to the second JMTR irradiation (70M-61P) of the demestic austenitic stainless steels for the structural material of the FBR performed by Power Reactor and Nuclear Fuel Development Corporation. The thermal control specimens were given the temperature history which simulated that of the irradiation temperature in vacuum by the electrical furnance, and then the tensile, fatigue and Charpy impact tests were performed. The changes of the material properties caused by the thermal history were investigated. (auth.)

Characterization of alternative FBR development strategies

International Nuclear Information System (INIS)

Boegel, A.J.; Clausen, M.J.

1981-08-01

Near-term decisions regarding the nature and place of the FBR development program must be made. This study is part of a larger program designed to provide the Department of Energy (DOE) with imformation that can be used to make strategic programmatic decisions. The focus of this report is the description of alternative approaches for developing the FBR and the quantification of the duration and cost of each alternative. The time frames of the alternative approaches are investigated in companion reports (White 1981 and Fraley 1981). The results of these analyses will be described in a summary report

Present state of development of demonstration FBR and prospect of practical use

International Nuclear Information System (INIS)

Inagaki, Tatsutoshi

1996-01-01

As for the FBR development in Japan, the Atomic Energy Commission revised the long term plan on the research, development and utilization of atomic energy in June, 1994, and under the basic policy that through the considerable period of using LWRs together, FBRs will be adopted as the main nuclear power plants in future, it was decided to establish FBR technology system so that the practical use of FBRs becomes feasible by about 2030 through two demonstration FBRs following the experimental FBR 'Joyo' and the prototype FBR 'Monju'. The Monju started power generation and transmission in August, 1995, but secondary sodium leak accident occurred in December, 1995, and at present it is stopped. The demonstration FBR No. 1 is a top entry type loop reactor, and the power output is about 660 MWe. The start of construction is scheduled at the beginning of 2000s. The research on the whole plant design is carried out as the research on the optimization of demonstration FBR plant for three years from fiscal year 1994. The design of the demonstration FBR No. 1, the research and development for it, the prospect of the practical use and the research and development for the practical use are reported. (K.I.)

Development of FBR cycle data base system (II)

International Nuclear Information System (INIS)

Kubota, Sadae; Ohtaki, Akira; Hirao, Kazuhiro

2003-05-01

In the 'Feasibility Study on Commercialized FBR Cycle Systems (F/S)', scenario evaluations, cost-benefit evaluations and system characteristic evaluations to show the significance of the FBR cycle system introduction concretely are performed while design studies for FBR plants, reprocessing systems and fabrication systems are conducted. In these evaluations, future society of various conditions and situation is assumed, and investigation and analysis about needs and social effects of FBR cycle are carried out. In this study, promising FBR cycle concepts are suggested by taking information such as domestic and foreign policies and bills, an economic prediction, a supply and demand prediction of resources, a project of technology development into consideration in addition to system design information. The development of the FBR Cycle Database which this report introduced started in 1999 fiscal year to enable managed unitarity and searched reference information to use for the above scenario evaluations, cost-benefit evaluations and system characteristic evaluations. In 2000 fiscal year, its prototype was made and used tentatively, and we extracted the problems in operation and functions from that, and, in 2001 fiscal year, the entry system and the search system using the Web page were made in order to solve problems of the prototype, and started use in our group. Moreover, in 2002 fiscal year, we expanded and improved the search system and promoted the efficiency of management work, and use in JNC through intranet of the database was started. In addition, as a result of having made the entry of about 350 data in 2002 fiscal year, the collected number of the database reaches about 7,250 by the end of March, 2003. We are to continue the entry of related information of various evaluations in F/S phase 2 from now on. In addition, we are to examine improvement of convenience of the search system and cooperation with the economy database. (author)

Progress in researches on MOX fuel pellet producing technology in China

International Nuclear Information System (INIS)

Hu Xiaodan

2010-01-01

Being the key section of nuclear-fuel cycle, the producing technology of MOX(UO 2 -PuO 2 ) fuel had driven to maturity in France, England, Russia, Belgium, etc. MOX fuel had been applied in FBR and LWR successfully in those countries. With the rapidly developing of nuclear-generated power, the MOX fuel for FBR and LWR was active demanded in China. However, the producing technology of MOX fuel developed slowly. During the period of 'the seventh five year's project', MOX fuel pellet was produced by mechanically mixed method and oxalate deposited method, respectively. Parts of cool performance of MOX fuel pellet produced by oxalate deposited method reached the qualification of fuel for FBR. During the period of 'the ninth five year's project' and 'the tenth five year's project', the technical route of producing MOX fuel was determined, and the test line of producing MOX fuel was built preliminarily. In the same time, the producing technology and analyzing technology of MOX fuel pellet by mechanically mixed was studied roundly, and the representative analogue pellet(UO 2 -CeO 2 ) was produced. That settled the supporting technology for the commercial process and research of MOX fuel rod and MOX fuel module. (authors)

Fuel assembly for FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki.

1995-01-01

Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

History and current status of nuclear fuel reprocessing technology

International Nuclear Information System (INIS)

Funasaka, Hideyuki; Nagai, Toshihisa; Washiya, Tadahiro

2008-01-01

History and present state of fast breeder reactor was reviewed in series. As a history and current status of nuclear fuel reprocessing technology, this ninth lecture presented the progress of the FBR fuel reprocessing technology and advanced reprocessing processes. FBR fuel reprocessing technology had been developed to construct the reprocessing equipment test facilities (RETF) based on PUREX process technologies. With economics, reduction of environmental burdens and proliferation resistance taken into consideration, advanced aqueous method for nuclear fuel cycle activities has been promoted as the government's basic policy. Innovative technologies on mechanical disassembly, continuous rotary dissolver, crystallizer, solvent extraction and actinides recovery have been mainly studied. (T. Tanaka)

Development of the tool for generating ORIGEN2 library based on JENDL-3.2 for FBR

International Nuclear Information System (INIS)

Ohkawachi, Yasushi; Fukushima, Manabu

1999-05-01

ORIGEN2 is one of the most widely-used burnup analysis code in the world. This code has one-grouped cross section libraries compiled for various types of reactors. However, these libraries have some problems. One is that these libraries were developed from old nuclear data libraries (ENDF/B-IV,V) and the other is that core and fuel designs from which these libraries are generated do not match the current analysis. In order to solve the problems, analysis tool is developed for generating ORIGEN2 library from JENDL-3.2 considering multi-energy neutron spectrum. And eight new libraries are prepared using this tool for analysis of sodium-cooled FBR. These new libraries are prepared for eight kinds of cores in total. Seven of them are made by changing core size (small core - large core), fuel type (oxide, nitride, metal) and Pu vector as a parameter. The eighth one is a Pu burner core. Burnup calculation using both new and original libraries, shows large difference in buildup or depletion numbers of nuclides among the libraries. It is estimated that the analysis result is greatly influenced by the neutron spectrum which is used in collapse of cross section. By using this tool or new libraries, it seems to improve evaluation accuracy of buildup or depletion numbers of nuclides in transmutation research on FBR fuel cycle. (author)

A study on adsorption onto TODGA resin after electrolytic reduction in ERIX process for reprocessing spent FBR-MOX fuel

International Nuclear Information System (INIS)

Hoshi, Harutaka; Arai, Tsuyoshi; Wei, Yuezhou; Kumagai, Mikio; Asakura, Toshihide; Morita, Yasuji

2005-01-01

For reprocessing spent FBR-MOX fuel, an advanced aqueous reprocessing process ''ERIX process'' has been developed. In this system, hydrazine is used as reduction holding reagent for the valance adjustment of U by electrolytic reduction in nitric acid solution. Therefore, hydrazine is contained in high level liquid waste after separation of U, Pu and Np. Effect of hydrazine on adsorption of FP onto TODGA resin was examined. When hydrazine concentration was less than 0.3 M, effect on the distribution coefficient was negligibly small. After electrolytic reduction, some elements exist as lower valence state. Ru and Tc are most difficult elements to control their behavior in aqueous process. The distribution coefficient of both Ru and Tc onto TODGA decreased after electrolytic reduction, because they are reduced to lower valence. Hence, it is difficult for Ru or Tc to diffuse to allover the process and separation of MA from Tc and Ru was enhanced by electrolytic reduction. (author)

Fuel exchange device for FBR type reactor

International Nuclear Information System (INIS)

Onuki, Koji.

1993-01-01

The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

Development of FBR cycle data base system

International Nuclear Information System (INIS)

Kubota, Sadae; Ohtaki, Akira; Hirao, Kazuhiro

2002-06-01

In the 'Feasibility Study on Commercialized Fast Reactor Cycle System (F/S)'. scenario evaluations, cost-benefit evaluations and system characteristic evaluations to show significance of the Fast Breeder Reactor (FBR) cycle system introduction concretely are performed in parallel with a design study for FBR plants, reprocessing systems and fabrication systems. In these evaluations, informations such as economic prospects, prospects for supply and demand of resources and a progress of engineering development are used in addition to design information. This report explains a FBR Cycle Database in order to carry out management and search of various design information and the relating information. The prototype system of the database was completed in the 2000 fiscal year, and the problem of the user number restriction of the prototype system has been improved by Web-ization in the 2001 fiscal year. About 7,000 data are stored in this data base (as of the end of March, 2002). The expansion of user etc., and the continuation of input work of various evaluation information will be carried out, in the phase 2 of F/S. (author)

Application of seismic isolation technology to demonstration FBR

International Nuclear Information System (INIS)

Kato, Muneaki

1994-01-01

The Japanese demonstration FBR is loop type, the intermediate heat exchanger is installed between the reactor and the steam generator, and up to the intermediate heat exchanger is in the containment vessel, which is designed as a reinforced concrete vessel. In FBRs, the optimization in aseismatic design and high temperature structural design is important. The reactor building is buried in rock bed up to its center of gravity to minimize the amplifying earthquake response. If the seismic isolation structure for a reactor is realized, cost reduction can be expected by the rationalization of machinery and equipment and the standardization of buildings and facilities. The research on FBR seismic isolation design has been carried out by Central Research Institute of Electric Power Industry and Japan Atomic Power Co. The concept of FBR seismic isolation design, the basic condition for the design evaluation, the research on safety allowance and the conceptual design analysis are reported. (K.I.)

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

Chinese FBR program and its first FR conceptual design

International Nuclear Information System (INIS)

Xu Mi

1991-01-01

As the preliminary study results, The long term strategy of FBR development in China would be divided into three steps: Experimental Fast Reactor, Modulized Fast Breeder Reactor and Large Fast Breeder Reactor. The design requirements, main technical selections and design boundary conditions for the first step, experimental fast reactor, which is named FFR (First Fast Reactor) in China have been given in the paper. The conceptual design of FFR, based on core neutronics, core subassemblies thermo-hydraulics, reactor block and mechanics, fuel handling system, main heat transfer system and steam-electricity generation system has been completed and also presented in this paper. Finally, the R and D program of FFR which is carring on and will be done is briefly introduced. (author)

CANDU fuel cycle options in Korea

International Nuclear Information System (INIS)

Boczar, P.G.; Fehrenbach, P.J.; Meneley, D.A.

1996-04-01

The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a twoto three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU 1 FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on

Application of the PSA method to decay heat removal systems in a large scale FBR design

International Nuclear Information System (INIS)

Kotake, S.; Satoh, K.; Matsumoto, H.; Sugawara, M.; Sakata, K.; Okabe, A.

1993-01-01

The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10 -7 /d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

Application of the PSA method to decay heat removal systems in a large scale FBR design

Energy Technology Data Exchange (ETDEWEB)

Kotake, S; Satoh, K [Japan Atomic Power Company, Otemachi, Chiyoda-ku, Tokyo (Japan); Matsumoto, H; Sugawara, M [Toshiba Corporation (Japan); Sakata, K [Mitsubishi Atomic Power Industries Inc. (Japan); Okabe, A [Hitachi Engineering Co., Ltd. (Japan)

1993-02-01

The Probabilistic Safety Assessment (PSA) method is applied to a large scale loop-type FBR in its conceptual design stage in order to establish a well-balanced safety. Both the reactor shut down and decay heat removal systems are designed to be highly reliable, e.g. 10{sup -7}/d. In this paper the results of several reliability analyses concerning the DHRS have been discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The reliability is evaluated small enough, since DRACSs consists of four independent loops with sufficient heat removal capacity and both forced and natural circulation capabilities are designed. It is found that the common mode failures for the active components in the DRACS dominate the reliability. The design diversity concerning these components can be effective for the improvements and the accident managements on BOP are also possible by making use of the long grace period in FBR. (author)

Development of a standard data base for FBR core nuclear design (XIII). Analysis of small sample reactivity experiments at ZPPR-9

International Nuclear Information System (INIS)

Sato, Wakaei; Fukushima, Manabu; Ishikawa, Makoto

2000-09-01

A comprehensive study to evaluate and accumulate the abundant results of fast reactor physics is now in progress at O-arai Engineering Center to improve analytical methods and prediction accuracy of nuclear design for large fast breeder cores such as future commercial FBRs. The present report summarizes the analytical results of sample reactivity experiments at ZPPR-9 core, which has not been evaluated by the latest analytical method yet. The intention of the work is to extend and further generalize the standard data base for FBR core nuclear design. The analytical results of the sample reactivity experiments (samples: PU-30, U-6, DU-6, SS-1 and B-1) at ZPPR-9 core in JUPITER series, with the latest nuclear data library JENDL-3.2 and the analytical method which was established by the JUPITER analysis, can be concluded as follows: The region-averaged final C/E values generally agreed with unity within 5% differences at the inner core region. However, the C/E values of every sample showed the radial space-dependency increasing from center to core edge, especially the discrepancy of B-1 was the largest by 10%. Next, the influence of the present analytical results for the ZPPR-9 sample reactivity to the cross-section adjustment was evaluated. The reference case was a unified cross-section set ADJ98 based on the recent JUPITER analysis. As a conclusion, the present analytical results have sufficient physical consistency with other JUPITER data, and possess qualification as a part of the standard data base for FBR nuclear design. (author)

Technological study report on synthetic evaluation for FBR cycle. The report of the feasibility studies on commercialized FBR cycle system. Phase 1

International Nuclear Information System (INIS)

Shinoda, Yoshihiko; Ohtaki, Akira; Kofuji, Hirohide; Ono, Kiyoshi; Hirao, Kazunori

2001-03-01

This report is intended to explain the outline of the characteristic evaluation work on various FR cycle system concepts, following the design work, in the 1st phase of the JNC's 'Feasibility Study on Commercialized Fast Reactor Cycle System (the F/S)' (from 1999 to March 2001). The purpose of this characteristic evaluation is to reveal the performance of candidate FR cycle systems. For this synthetic estimation, six viewpoints, such as Economics, Effective utilization of uranium resource, Reduction of environmental impact, Safety, Proliferation resistance, and Technological feasibility, are selected. In addition, aiming at the practical use in phase 2, we examined an application to FBR research and development of cost benefit analysis method used for the policy evaluation. Furthermore, long-term nuclear material mass flow was analyzed and the scenario of 'FBR application for the hydrogen production' is proposed, considering how FBR would be utilized for the 21st century. And, a database including the various documents and data used for evaluation was constructed. (author)

Flexible fuel cycle initiative for the transition period from current reactors to next generation reactors

International Nuclear Information System (INIS)

Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi; Kawamura, Fumio; Shiina, Kouji; Sasahira, Akira

2005-01-01

A sustainable electricity supply by fast breeder reactors (FBRs) is essential to ensure energy security and prevent global warming. Transition from light water reactors (LWRs) to FBRs and establishment of an FBR cycle are indispensable, which requires plutonium (Pu) for the introduction of FBRs. The authors propose advanced system called 'Flexible Fuel Cycle Initiative (FFCI)' which can respond flexibly the future expected technical and social uncertainties, can hold no surplus Pu, and can achieve an economical FBR cycle. In the new concept of FFCI, 2nd LWR reprocessing which would succeed Rokkasho Reprocessing Plant is a simple facility to carry out only uranium (U) removal and residual 'recycle material' is stored or utilized. According to FBRs introduction status, recycle material is immediately treated in an FBR reprocessing to fabricate FBR fuel or temporarily stored for the utilization in FBRs at necessary timing. FFCI has high flexibility by having several options for future uncertainties by the introduction of recycle material as a buffer material between LWR and FBR cycles. (author)

Development of flaw assesment methodology for elevated temperature components of FBR plants

International Nuclear Information System (INIS)

Shimakawa, Takashi; Takahashi, Yukio; Miura, Naoki; Nakayama, Yasunari; Sawai, Tatsuaki; Tooya, Yuuji

1999-01-01

Fracture mechanics is applicable for the safety assessment of FBR component if a crack is assumed to exist. Inelastic response should be taken into account due to high temperature operation of FBR components. However, methodology for the application of inelastic fracture mechanics has not been established sufficiently. CRIEPI has been conducted research projects to develop a flaw assessment guideline for FBR components. This guideline consists of evaluation methods for creep-fatigue crack propagation, ductile fracture and sodium leak rate. The summary of evaluation methods on creep-fatigue crack and ductile fracture is presented in this paper. (author)

Fuel assemblies for FBR type reactor

International Nuclear Information System (INIS)

Ikeda, Kiyoshi.

1981-01-01

Purpose: To decrease errors in the flow rate distribution of coolants by resiliently inserting a flow regulation rod having a variable flow regulation element formed at the upper portion along the axial direction in the entrance nozzle of a fuel assembly. Constitution: A plurality of orifice aperture are formed to the entrance nozzle of a fuel assembly and an aperture for inserting a flow regulation rod is formed to the top end of the entrance nozzle. A fixed flow regulation element A and a variable flow regulation element B supported coaxially with the nozzle by a support ring are disposed to the inside of the nozzle. The element B is urged by the resilient urging spring to the element A and connected by way of support lever to the flow regulation rod. While on the other hand, the top end of the nozzle is inserted through the partition wall between a high pressure coolant chamber and a low pressure coolant chamber. An aperture for hydrodynamically supporting the fuel assembly is provided by way of a frame and a flow regulation rod that stands vertically from the low pressure coolant chamber is disposed to the center of the frame. In the fuel assembly, the flow regulation rod inserted from the aperture at the top end of the nozzle pushes the element B upwardly to thereby maintain a flow passage of the coolant between the elements A and B. (Seki, T.)

Accurate reactivity void coefficient calculation for the fast spectrum reactor FBR-IME

Energy Technology Data Exchange (ETDEWEB)

Lima, Fabiano P.C.; Vellozo, Sergio de O.; Velozo, Marta J., E-mail: fabianopetruceli@outlook.com, E-mail: vellozo@cbpf.br, E-mail: martajann@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Militar

2017-07-01

This paper aims to present an accurate calculation of the void reactivity coefficient for the FBR-IME, a fast spectrum reactor in development at the Engineering Military Institute (IME). The main design peculiarity lies in using mixed oxide [MOX - PuO{sub 2} + U(natural uranium)O{sub 2}] as fuel core. For this task, SCALE system was used to calculate the reactivity for several voids distributions generated by bubbles in the sodium beyond its boiling point. The results show that although the void reactivity coefficient is positive and location dependent, they are offset by other feedback effects, resulting in a negative overall coefficient. (author)

Development of advanced methodology for defect assessment in FBR power plants

International Nuclear Information System (INIS)

Meshii, Toshiyuki; Asayama, Tai

2001-03-01

As a preparation for developing a code for FBR post construction code, (a) JSME Code NA1-2000 was reviewed on the standpoint of applying it to FBR power plants and the necessary methodologies for defect assessment for FBR plants were pointed out (b) large capacity-high speed fatigue crack propagation (FCP) testing system was developed and some data were acquired to evaluate the FCP characteristics under thermal stresses. Results showed that the extended research on the following items are necessary for developing FBR post construction code. (1) Development of assessment for multiple defects due to creep damage. Multiple defects due to creep damage are not considered in the existing code, which is established for nuclear power plants in service under negligible-creep temperature. Therefore method to assess the integrity of these multiple defects due to creep damage is necessary. (2) FCP resistance for small load. Since components of FBR power plants are designed to minimize thermal stresses, the accuracy of FCP resistance for small load is important to estimate the crack propagation under thermal stresses accurately. However, there is not a sufficient necessary FCP data for small loads, maybe because the data is time consuming. Therefore we developed a large capacity-high speed FCP testing system, made a guideline for accelerated test and acquired some data to meet the needs. Continuous efforts to accumulate small load FCP data for various materials are necessary. (author)

Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

International Nuclear Information System (INIS)

Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

2010-01-01

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

Energy Technology Data Exchange (ETDEWEB)

Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

1999-07-01

The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

Analytical methodology and facility description spent fuel policy

Energy Technology Data Exchange (ETDEWEB)

1978-08-01

Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs.

Analytical methodology and facility description spent fuel policy

International Nuclear Information System (INIS)

1978-08-01

Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs

Continuous Ethanol Production Using Immobilized-Cell/Enzyme Biocatalysts in Fluidized-Bed Bioreactor (FBR)

Energy Technology Data Exchange (ETDEWEB)

Nghiem, NP

2003-11-16

The immobilized-cell fluidized-bed bioreactor (FBR) was developed at Oak Ridge National Laboratory (ORNL). Previous studies at ORNL using immobilized Zymomonas mobilis in FBR at both laboratory and demonstration scale (4-in-ID by 20-ft-tall) have shown that the system was more than 50 times as productive as industrial benchmarks (batch and fed-batch free cell fermentations for ethanol production from glucose). Economic analysis showed that a continuous process employing the FBR technology to produce ethanol from corn-derived glucose would offer savings of three to six cents per gallon of ethanol compared to a typical batch process. The application of the FBR technology for ethanol production was extended to investigate more complex feedstocks, which included starch and lignocellulosic-derived mixed sugars. Economic analysis and mathematical modeling of the reactor were included in the investigation. This report summarizes the results of these extensive studies.

Evaluation of effects of different fuel cycles in environment

International Nuclear Information System (INIS)

1979-03-01

The paper discusses the risks for personnel and environment caused by different steps of FBR-Fuel cycle. For FBRS the most important factor of exposure is the fabrication of fuel elements but technical development is capable to reduce this factor

Temperature fluctuation reducing device for FBR type reactor

International Nuclear Information System (INIS)

Ootsuka, Fumio; Shiratori, Fumihiro.

1991-01-01

In existent FBR type reactors, since temperature fluctuation in the reactor upper portion has been inevitable, thermal fatigue may be caused possibly in reactor core upper mechanisms. Then, a valve is disposed to a control rod lower guide tube contained in a reactor container for automatically controlling the amount of passing coolants in accordance with the temperature of the passing coolants, to mix and control coolants passing through a fuel assembly in adjacent with the guide tube and coolants passing through the guide tube. Further, a rectification cylinder is disposed, in which a portion of coolants passing through the fuel assembly is caused to flow. An orifice is disposed to the cylinder with an exit being disposed to the upstream thereof such that the coolants not flown into the rectification cylinder and the coolants passing through the guide tube are mixed to moderate the temperature fluctuation. That is, a portion of the coolants flown into the rectification cylinder can not pass through the orifice, but flow backwardly to the upstream and is discharged out of the rectification cylinder from the coolants exit and mixed sufficiently with coolants passing through the guide tube. In this way, temperature fluctuation can be moderated. (N.H.)

A review on manufacturing technology for long-lived radionuclide fuel compounds

International Nuclear Information System (INIS)

Hwang, Doo Seong; Park, Jin Ho; Kim, Eung Ho; Chung, Won Myung; Lee, Kui Ill; Woo, Moon Sik; Kim, Yeon Ku; Yoo, Jae Hyung

1998-03-01

Thermal neutron reactor (LWR), fast neutron reactor (FBR), accelerator-driven subcritical system have been studied as the potential transmutation devices. The fuel types can be classified according to the concept of each reactor. Oxide fuel is considered in LWR and metal, oxide, and nitride fuels are studied in FBR. In accelerator-driven subcritical system molten salt, metal, and oxide fuels are considered. This review focused on characteristics according to transmutation system, and manufacturing technologies of each fuels. Accelerator-driven system is being proposed as the most reasonable concept in recent, since it has merits in terms of stability and free control of nuclides composition rate in charge of long-lived nuclides. Fluorides molten salt fuel is better chemically stable and corrosion resistant, and lower vapor pressure than chloride molten salt and metal in the fuel type of accelerator-driven system. And then the detail manufacturing technology of fluorides molten salt were reviewed. (author). 62 refs., 23 tabs., 37 figs

Heavy liquid metal cooled FBR. Results 2001

International Nuclear Information System (INIS)

Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu

2003-08-01

In the feasibility studies of commercialization of an FBR fuel cycle system, the targets are economical competitiveness to future LWRs, efficient utilization of resources, reduction of environmental burden and enhancement of nuclear non-proliferation, besides ensuring safety. Both medium size pool-type lead-bismuth cooled reactor with primary pumps system and without primary pumps system are studied to pursue their improvement in heavy metal coolant considering design requirements form plant structures. The design of plant systems are reformed, and the conceptual design is made and the commodities are analyzed. (1) Conceptual design of lead-bismuth cooled reactor with pumping system: Electrical output 750 MWe and 4-module system. The heat-mass balance is optimized and drawings are made about plant layout, cooling system, reactor structure and cooling component structures. (2) Structural analysis of main components. (3) Conceptual design of natural circulation type lead-bismuth cooled reactor: Electrical output 550 MWe and 6-module system. The heat-mass balance is optimized and drawings are made about plant layout, cooling system, reactor structure and cooling component structures. (4) Study of R and D program. (author)

Two-dimensional analytical model of a proton exchange membrane fuel cell

International Nuclear Information System (INIS)

Liu, Jia Xing; Guo, Hang; Ye, Fang; Ma, Chong Fang

2017-01-01

In this study, a two-dimensional full cell analytical model of a proton exchange membrane fuel cell is developed. The analytical model describes electrochemical reactions on the anode and cathode catalyst layer, reactants diffusion in the gas diffusion layer, and gases flow in the gas channel, etc. The analytical solution is derived according to the basic physical equations. The performance predicted by the model is in good agreement with the experimental data. The results show that the polarization mainly occurs in the cathode side of the proton exchange membrane fuel cell. The anodic overpotential cannot be neglected. The hydrogen and oxygen concentrations decrease along the channel flow direction. The hydrogen and oxygen concentrations in the catalyst layer decrease with the current density. As predicted by the model, concentration polarization mainly occurs in the cathode side. - Highlights: • A 2D full cell analytical model of a proton exchange membrane fuel cell is developed. • The analytical solution is deduced according to the basic equations. • The anode overpotential is not so small that it cannot be neglected. • Species concentration distributions in the fuel cell is obtained and analyzed.

Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

International Nuclear Information System (INIS)

Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

1999-07-01

A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

A novel analytical analysis of PEM fuel cell

International Nuclear Information System (INIS)

Yazdi, Mohamad Zardoshtizade; Kalbasi, Mansour

2010-01-01

In this study, a quasi three-dimensional model was developed for a single proton exchange membrane (PEM) fuel cell. The model was used for a half-cell which includes the cathode gas channel, gas diffusion layer (GDL), cathode catalyst layer and membrane is modeled. This model includes mass transfer in the gas channel and GDL, electrochemistry reaction in cathode catalyst layer, and charge transfer in the membrane phase. These expressions were solved by analytical methods. An agglomerate approach was used to describe cathode catalyst layer. By using analytical solution, the expressions can predict the PEM fuel cell behavior in different conditions which is the advantage of this method.

Analytic tests and their relation to jet fuel thermal stability

Energy Technology Data Exchange (ETDEWEB)

Heneghan, S.P.; Kauffman, R.E. [Univ. of Dayton Research Institute, OH (United States)

1995-05-01

The evaluation of jet fuel thermal stability (TS) by simple analytic procedures has long been a goal of fuels chemists. The reason is obvious: if the analytic chemist can determine which types of material cause his test to respond, the refiners will know which materials to remove to improve stability. Complicating this quest is the lack of an acceptable quantitative TS test with which to compare any analytic procedures. To circumvent this problem, we recently compiled the results of TS tests for 12 fuels using six separate test procedures. The results covering a range of flow and temperature conditions show that TS is not as dependent on test conditions as previously thought. Also, comparing the results from these tests with several analytic procedures shows that either a measure of the number of phenols or the total sulfur present in jet fuels is strongly indicative of the TS. The phenols have been measured using a cyclic voltammetry technique and the polar material by gas chromatography (atomic emission detection) following a solid phase extraction on silica gel. The polar material has been identified as mainly phenols (by mass spectrometry identification). Measures of the total acid number or peroxide concentration have little correlation with TS.

Analytical prediction of fuel assembly spacer grid loss coefficient

International Nuclear Information System (INIS)

Lim, J. S.; Nam, K. I.; Park, S. K.; Kwon, J. T.; Park, W. J.

2002-01-01

The analytical prediction model of the fuel assembly spacer grid pressure loss coefficient has been studied. The pressure loss of gap between the test section wall and spacer grid was separated from the current model and the different friction drag coefficient on spacer straps from high Reynolds number region were used to low Reynolds number region. The analytical model has been verified based on the hydraulic pressure drop test results for the spacer grids of three types for 5x5, 16x16(or 17x17) arrays. The analytical model predicts the pressure loss coefficients obtained from test results within the maximum errors of 12% and 7% for 5x5 test bundle and full size bundle, respectively, at Reynolds number 500,000 of the core operating condition. This result shows that the analytical model can be used for research and design change of the nuclear fuel assembly

Long-term logistic analysis of FBR introduction strategy: avoiding both uranium and plutonium shortage

International Nuclear Information System (INIS)

Suzuki, T.

1995-01-01

Despite comfortable predictions on short to mid-term uranium resources, there is still a concern about long-term availability of competitive uranium resources. In order to achieve substantial uranium saving, early introduction of Fast Breeder Reactor (FBR) is desirable. But it is also known that rapid introduction of FBR could result in plutonium storage. Will there be enough plutonium on a global scale to sustain fast FBR growth? is there any other way to save uranium resource? This paper concludes that multi-option strategies to achieve flexible long-term strategy to avoid both uranium and plutonium storage are desirable. (authors)

Melting temperature of uranium - plutonium mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Ishii, Tetsuya; Hirosawa, Takashi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1997-08-01

Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960`s and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960`s and that some of the 1960`s data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO{sub 2} - PuO{sub 2} - PuO{sub 1.61} ideal solution model, and then formulized. (J.P.N.)

Melting temperature of uranium - plutonium mixed oxide fuel

International Nuclear Information System (INIS)

Ishii, Tetsuya; Hirosawa, Takashi

1997-08-01

Fuel melting temperature is one of the major thermodynamical properties that is used for determining the design criteria on fuel temperature during irradiation in FBR. In general, it is necessary to evaluate the correlation of fuel melting temperature to confirm that the fuel temperature must be kept below the fuel melting temperature during irradiation at any conditions. The correlations of the melting temperature of uranium-plutonium mixed oxide (MOX) fuel, typical FBR fuel, used to be estimated and formulized based on the measured values reported in 1960's and has been applied to the design. At present, some experiments have been accumulated with improved experimental techniques. And it reveals that the recent measured melting temperatures does not agree well to the data reported in 1960's and that some of the 1960's data should be modified by taking into account of the recent measurements. In this study, the experience of melting temperature up to now are summarized and evaluated in order to make the fuel pin design more reliable. The effect of plutonium content, oxygen to metal ratio and burnup on MOX fuel melting was examined based on the recent data under the UO 2 - PuO 2 - PuO 1.61 ideal solution model, and then formulized. (J.P.N.)

Development of test apparatus for fission product release from overheated fuel element

International Nuclear Information System (INIS)

Takai, T.; Hirosawa, T.; Funabashi, H.; Miyahara, S.

1996-01-01

Evaluation of the source term released to environment under the accident conditions is important to the safety evaluation and design of reactor containment vessel. However, available data related to FBR source term are very limited, especially for the FPs release data from overheated FBR fuel. The present, source term evaluation of FBR is based on assumption from that of LWR. Though, this evaluation is very conservative. Evaluation large scale FBR source term using this method is result in extremely conservative and lead construction of large scale plant becomes doubtful from the viewpoints of cost and safety system. Though, it is necessary to evaluate source term from the realistic and rational scenario considering a characteristic of FBR. Preparation of FPs release experiment from irradiated fuel is going on to investigate the FPs release and transport and to develop the analysis code for in-vessel source term evaluation. Fabrication of this apparatus was started in 1992, and the installation was completed in 1994. This apparatus passed the facility inspection by Science and Technology Agency in March 1995. This apparatus consists of a high frequency induction furnace, thermal gradient tube (TGT), sintered metal filters, cold traps, gas-analyzer, γ-ray spectrometry system and so on. In the experiment, FPs release rate and behavior will be investigated using gamma-ray spectrogram and FP gas analysis. Physical and chemical composition of released FP would be investigated from FPs deposited profiles on TGT. Now, cold experiment using simulant FP materials are conducted. (author)

CANDU fuel cycle options in Korea

International Nuclear Information System (INIS)

Boczar, P. G.; Fehrenbach, P. J.; Meneley, D. A.

1996-01-01

FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU/FBR system, in which the FBR would be operated as a 'fuel factory,'providing the fissile material to power a number of lower-cost, high-efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on-line fuelling provide flexibility to respond to changing fuel-cycle requirements in the short term and in the indefinite future

Current status of feasibility studies on commercialized fuel cycle system for Fast Breeder Reactor

International Nuclear Information System (INIS)

Ojima, Hisao; Nagaoki, Yoshihiro

2000-01-01

A 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' is underway at the Japan Nuclear Cycle Development Institute (JNC). The study will select the promising concepts with their R and D tasks in order to commercialize the fast breeder reactor (FBR) cycle system. The feasibility studies (F/S) have to present surveyed and screened various relevant technologies, and defined the design requirement of the commercialized fuel cycle system for FBR. The promising technical options are being evaluated and conceptual designs are being examined. At the end of JFY2000, several candidate concepts of the commercialized FBR cycle system will be proposed. (author)

Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

International Nuclear Information System (INIS)

Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

2000-01-01

The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

Outline of the seismic design guideline of an FBR - a tentative draft

International Nuclear Information System (INIS)

Akiyama, Hiroshi; Ohtsubo, Hideomi; Nakamura, Hideharu; Matsuura, Shinichi; Hagiwara, Yutaka; Yuhara, Tetsuo; Hirayama, Hiroshi; Kokubo, Kunio; Ooka, Yuji.

1993-01-01

Central Research Institute of Electric Power Industry (Japan) is carrying out the Demonstration Test and Research Program of Buckling of FBR (FY 1987-FY 1993). The first half of the research program was finished after establishing a seismic buckling design guideline (a tentative draft). The purpose of this paper is to describe the dynamic buckling characteristics of FBR main vessels and the outline of the rationalized buckling design guideline for seismic loadings. (orig.)

Fuel assembly

International Nuclear Information System (INIS)

Bando, Masaru.

1993-01-01

As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

Fuel assemblies

International Nuclear Information System (INIS)

Nakatsuka, Masafumi.

1979-01-01

Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

Method for selecting FBR development strategies in the presence of uncertainty

International Nuclear Information System (INIS)

Fraley, D.W.; Burnham, J.B.

1981-12-01

This report describes the methods used to probabilistically analyze data related to the uranium supply the FBR's competitive dates, development strategies' time and costs, and economic benefits. It also describes the econometric methods used to calculate the economic risks of mistiming the development. Seven strategies for developing the FBR are analyzed. The various measures of a strategy's performance - timing, costs, benefits, and risks - are combined into several criteria which are used to evaluate the seven strategies. Methods are described for selecting a strategy based on a number of alternative criteria

Treatment of PAHs in waters using the GAC-FBR process

International Nuclear Information System (INIS)

Hickey, R.F.; Sunday, A.; Wagner, D.; Groshko, V.; Rajan, R.V.; Leuschner, A.; Hayes, T.D.

1995-01-01

Pilot studies were conducted to determine the utility of the granular activated carbon fluidized-bed reactor (GAC-FBR) process to treat groundwater from manufactured gas plant (MGP) sites containing polycyclic aromatic hydrocarbons (PAHs) and a process effluent water from a deep subsurface dense, nonaqueous-phase liquid (DNAPL) removal process at an MGP site. Removal of naphthalene exceeded 99.9%, and overall PAH removals of 99+% were observed at organic loading rates (OLRs) exceeding 4-kg chemical oxygen demand (COD)/m 3 -d and a hydraulic retention time (HRT) of about 6 min. Analysis of PAHs accumulated on GAC and oxygen consumption clearly demonstrated that removal of 2- to 4-ring PAHs was due primarily to biological oxidation and not to adsorption. Analysis of influent and effluent samples using Microtox reg-sign indicated removal of toxicity. Full-scale application of the GAC-FBR process has begun at a Superfund site in Pennsylvania. The GAC-FBR is being used to treat a 15-gal per min (gpm) process effluent flow from a subsurface DNAPL removal process. Initial results confirm the ability of the process to treat PAHs at high OLRs and short HRTs

The manual of a computer software 'FBR Plant Planning Design Prototype System'

International Nuclear Information System (INIS)

2003-10-01

This is a manual of a computer software 'FBR Plant Planning Design Prototype System', which enables users to conduct case studies of deviated FBR design concepts based on 'MONJU'. The calculations simply proceed as the user clicks displayed buttons, therefore step-by-step explanation is supposed not be necessary. The following pages introduce only particular features of this software, i.e, each interactive screens, functions of buttons and consequences after clicks, and the quitting procedure. (author)

The questions of liquid metal two-phase flow modelling in the FBR core channels

International Nuclear Information System (INIS)

Martsiniouk, D.Ye.; Sorokin, A.P.

2000-01-01

The two-fluid model representation for calculations of two-phase flow characteristics in the FBR fuel pin bundles with liquid metal cooling is presented and analysed. Two conservation equations systems of the mass, momentum and energy have been written for each phase. Components accounted the mass-, momentum- and heat transfer throughout the interface occur in the macro-field equations after the averaging procedure realisation. The pattern map and correlations for two-fluid model in vertical liquid metal flows are presented. The description of processes interphase mass- and heat exchange and interphase friction is determined by the two-phase flow regime. The opportunity of the liquid metal two-phase flow regime definition is analysed. (author)

Fuel assemblies for use in FBR type reactors

International Nuclear Information System (INIS)

Nakahigashi, Shigeo; Terasawa, Michitaka.

1984-01-01

Purpose: To prevent slackings in lapping wires and thereby enabling to maintain the distance between fuel pins always constant during use. Constitution: Lapping wires are wound helically around the outer circumference of each fuel pin in order to maintain the distance between fuel pins constant and unify the flow of coolants. The material of the lapping wire is defined as below. Specifically, austenite stainless steels incorporated with 0.045% titanium are used in the state of molten procession material as they are without no further cold working. Lapping wires having anti-swelling property can be obtained with this material and the slackings in the lapping wires during use can be prevented. (Ikeda, J.)

Defect analysis of BaSrFBr:Eu irradiated by X-ray

International Nuclear Information System (INIS)

Lee, C. Y.; Jeong, J. M.; Kim, J. H.

2010-01-01

The mechanical property of the BaSrFBr:Eu phosphor layer of X-ray image plates was investigated by using image quality (IQ), resolution (LP/mm), and coincidence Doppler broadening (CDB) positron annihilation. The screen samples of BaSrFBr:Eu phosphors were irradiated with hospital X-rays in the course of diagnostic radiography at an average rate of 20,000 times per year and were used for various periods of time. The LP/mm values of the irradiated BaSrFBr:Eu image plates varied between 2.4 and 2.0 for three years while the IQ values varied between 35 and 11 over the same period. CDB positron annihilation spectroscopy was used to analyze the defect structures in the phosphor layer. The S parameter values increased in correlation with increased exposure time, which indicated that more defects were generated. There was a positive relationship between the IQ and S parameters. Measurements of the defects indicate that most of the defects were likely to have been generated by the X-ray radiation.

Multiple recycling of fuel in prototype fast breeder reactor

Indian Academy of Sciences (India)

In the FBR closed fuel cycle, possibility of multi-recycle has been recognized. In the present study, Pu-239 equivalence approach is used to demonstrate the feasibility of achieving near constant input inventory of Pu and near stable Pu isotopic composition after a few recycles of the same fuel of the prototype fast breeder ...

Part 5. Fuel cycle options

International Nuclear Information System (INIS)

Lineberry, M.J.; McFarlane, H.F.; Amundson, P.I.; Goin, R.W.; Webster, D.S.

1980-01-01

The results of the FBR fuel cycle study that supported US contributions to the INFCE are presented. Fuel cycle technology is reviewed from both generic and historical standpoints. Technology requirements are developed within the framework of three deployment scenarios: the reference international, the secured area, and the integral cycle. Reprocessing, fabrication, waste handling, transportation, and safeguards are discussed for each deployment scenario. Fuel cycle modifications designed to increase proliferation defenses are described and assessed for effectiveness and technology feasibility. The present status of fuel cycle technology is reviewed and key issues that require resolution are identified

Uranium and plutonium distribution in unirradiated mixed oxide fuel from industrial fabrication

International Nuclear Information System (INIS)

Hanus, D.; Kleykamp, H.

1982-01-01

Different process variants developed in the last few years by the firm ALKEM to manufacture FBR and LWR mixed oxide fuel are given. The uranium and plutonium distribution is determined on the pellets manufactured with the help of the electron beam microprobe. The stepwise improvement of the uranium-plutonium homogeneity in the short-term developed granulate variants and in the long-term developed new processes are illustrated starting with early standard processes for FBR fuel. An almost uniform uranium-plutonium distribution could be achieved for the long-term developed new processes (OKOM, AuPuC). The uranium-plutonium homogeneity are quantified in the pellets manufactured according to the considered process variants with a newly defined quality number. (orig.)

Comparisons among different development ways of advanced reactors in China

International Nuclear Information System (INIS)

Guo Xingqu; Lin Jianwen; Wang Ruoli

1992-03-01

For the development of nuclear energy in the 21st century, China will select a new type reactor to develop, which will have higher fuel efficiency, high safety and better economics. The selection is among the types of FBR (fast breeder reactor), HTGR (high temperature gas-cooled reactor) and FFHR (fusion-fission hybrid reactor). Since the evaluation of advanced reactors involves many uncertain factors and the difficulty of quantization, both the AHP (analytic hierarchy process) method and expert consultation are adopted. Four aspects are taken in the norm system of AHP, i.e. safety, maturity of technology, economy and appropriateness. By using questionnaire method to experts and studying related documents, five types of advanced reactor are selected, i.e. oxide fueled FBR, metal fueled FBR, uranium fueled HTGR, U-Th fueled HTGR and FFBR. Their evaluation parameters are a comprehensively assessed and sorted. About 130 experts and professors who have been working in the research institutes and government agencies of nuclear field are asked to give their comments on the development of advanced reactors. The response rate of questionnaires is 86%, and the data collected are processed by computers. From the evaluation result of AHP method and expert consultation of the fast breeder reactor, especially, the metal fueled FBR, should have the priority in nuclear energy development in the 21st century in China

Method of locating a leaking fuel element in a fast breeder power reactor

International Nuclear Information System (INIS)

Honekamp, J.R.; Fryer, R.M.

1976-01-01

Failed fuel detection and location in an FBR is accomplished by mass spectrometric analysis of the cover gas for the 134 Xe/ 133 Xe ratio, and correlation with the theoretical ratio in fuel elements of known power level and burnup. (E.C.B.)

Control rod drives for FBR type reactor

International Nuclear Information System (INIS)

Ikakura, Hiroaki.

1990-01-01

The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

Current status of PIE activities in O-arai Engineering Center of JNC on FBR MOX fuel

International Nuclear Information System (INIS)

Koyama, Shin-ichi; Osaka, Masahiko; Namekawa, Takashi; Itoh, Masahiko

2003-01-01

Japan Nuclear Cycle Development Institute (JNC) is now totally promoting the development of commercialized fast reactors to realize stable supply of energy in future. One of the important items is to develop high-performance fuel. For this purpose, it is essential to carry out post-irradiation examinations (PIE) for evaluation of irradiated fuel performance and also to establish the PIE technology. This paper describes the current status of PIE results including its technology in O-arai Engineering Center of JNC. The facilities have been operating safely and successfully since the 1960's. Obtained PIE data were reflected to the design and operation of the experimental fast reactor JOYO, the prototype fast reactor MONJU and future fast reactors. The core modification from the breeding core (MK-I) to the irradiation core (MK-II) of JOYO was performed in 1982. Irradiation tests of fuels and materials in MK-II core started in 1982. At PIE facilities in OEC, 65 of driver fuels, fuel irradiation test rigs, material irradiation test rigs and several other components were examined related to JOYO MK-II core operation, and thus a lot of aspects were accumulated for irradiated fuel behaviors. As topical activities of these PIE techniques, burnup measurement and analytical technique for Minor Actinides (MA), such as neptunium and americium were described here. (author)

Analytical chemistry needs for nuclear safeguards in nuclear fuel reprocessing

International Nuclear Information System (INIS)

Hakkila, E.A.

1977-01-01

A fuel reprocessing plant designed to process 1500 tons of light water reactor fuel per year will recover 15 tons of Pu during that time, or approximately 40 to 50 kg of Pu per day. Conventional nuclear safeguards accountability has relied on batch accounting at the head and tail ends of the reprocessing plant with semi-annual plant cleanout to determine in-process holdup. An alternative proposed safeguards system relies on dynamic material accounting whereby in-line NDA and conventional analytical techniques provide indications on a daily basis of SNM transfers into the system and information of Pu holdup within the system. Some of the analytical requirements and problems for dynamic materials accounting in a nuclear fuel reprocessing plant are described. Some suggestions for further development will be proposed

Annual report of Power Reactor and Nuclear Fuel Development Corporation, fiscal year 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

The experimental FBR `Joyo` has continued the irradiation operation at 100 MWt. After the 11th periodic inspection, the 30th cycle operation was carried out. The cumulative operation time as of the end of the fiscal year was 51,630 hours, and the cumulative heat output was about 4.2 billion kWh. The prototype FBR `Monju` has succeeded in electric power generation in August, 1995, but the sodium leak accident occurred in December, 1995. The elucidation of the cause of the sodium leak accident and the total inspection for the safety have been carried out. As for FBRs, the research and development of the reactor physics, the design of a large FBR, the equipment systems, the fuel and materials, the structures and the safety have been advanced. The ATR `Fugen` Power Station has continued the operation smoothly, and as of the end of the fiscal year, the total generated electric power was about 17.3 billion kWh, and the capacity factor was 66.3%. It boasts about the result of using MOX fuel. The exploration of uranium resources, the development of uranium conversion, uranium enrichment and plutonium fuel, the reprocessing of spent fuel, the development of environmental technology for radioactive waste, creative and innovative research and development, safety control and safety research and others are reported. (K.I.)

Early-in-life thermal performance of UO2--PuO2 fast reactor fuel

International Nuclear Information System (INIS)

Baker, R.B.; Leggett, R.D.

1979-01-01

Results from the combined analyses of two thermal performance tests, HEDL P-19 and HEDL P-20 are described. The tests were designed to provide data on the power required to cause incipient fuel melting early in life under conditions prototypic of FFTF driver fuel pins and similar FBR fuel systems

Fabrication of Metallic Fuel Slugs for Irradiation Experiments in Fast Breeder Test Reactor

International Nuclear Information System (INIS)

Saify, M.T.; Jha, S.K.; Abdulla, K.K.; Kumar, Arun; Prasad, G.J.

2013-01-01

Advantages of Metallic fuels for future FBR: → High heavy metal atom density; → Higher thermal conductivity at room temperature that increases with temperature; → Metal fuels can be relatively easily fabricated with close dimensional tolerances; → They have excellent compatibility with liquid metal coolants

Analytical study of stress and deformation of HTR fuel blocks

International Nuclear Information System (INIS)

Tanaka, M.

1982-01-01

A two-dimensional finite element computer code named HANS-GR has been developed to predict the mechanical behavior of the graphite fuel blocks with realistic material properties and core environment. When graphite material is exposed to high temperature and fast neutron flux of high density, strains arise due to thermal expansion, irradiation-induced shrinkage and creep. Thus stresses and distortions are induced in the fuel block in which there are spatial variation of these strains. The analytical method used in the program to predcit these induced stresses and distortions by finite element method is discussed. In order to illustrate the versatility of the computer code, numerical results of two example analyses of the multi-hole type fuel elements in the VHTR Reactor are given. Two example analyses presented are those concerning the stresses in fuel blocks with control rod holes and distortions of the fuel blocks at the periphery of the reactor core. It is considered these phenomena should be carefully examined when the multi-hole type fuel elements are applied to VHTR. It is assured that the predicted mechanical behavior of the graphite components is strongly dependent on the material properties used and obtaining the reliable material property is important to make the analytical prediction a reliable one

JOYO modification program for demonstration tests of FBR innovative technology development

International Nuclear Information System (INIS)

Yoshimi, H.; Hachiya, Y.

1990-01-01

A plan is under way at PNC to modify the experimental fast reactor JOYO. The project is called MARK-III (MK-III) program. The purpose of MK-III is to expand the function of JOYO, and to make it possible to receive demonstration tests of new or high level technologies for FBR development. The MK-III program consists of two main modifications: conversion to a highly efficient irradiation facility; and a modification for demonstration testing of new technologies and concepts that have a high potential to reduce FBR plant construction cost, to evaluate plant reliability and to improve plant safety. These modifications are scheduled to start in 1991

Study on the nitride fuel fabrication for FBR cycle (1)

International Nuclear Information System (INIS)

Shinkai, Yasuo; Ono, Kiyoshi; Tanaka, Kenya

2002-07-01

In the phase-II of JNC's 'Feasibility Study on Commercialized Fuel Reactor Cycle System (the F/S)', the nitride fuels are selected as candidate for fuels for heavy metal cooled reactor, gas cooled reactor, and small scale reactor. In particular, the coated fuel particles are a promising concept for gas cooled reactor. In addition, it is necessary to study in detail the application possibility of pellet nitride fuel and vibration compaction nitride fuel for heavy metal cooled reactor and small scale reactor in the phase-II. In 2001, we studied more about additional equipments for the nitride fuel fabrication in processes from gelation to carbothermic reduction in the vibration compaction method. The result of reevaluation of off-gas mass flow around carbothermic reduction equipment in the palletizing method, showed that quantity of off-gas flow reduced and its reduction led the operation cost to decrease. We studied the possibility of fabrication of large size particles in the coated fuel particles for helium gas cooled reactor and we made basic technical issues clear. (author)

The current status of research and development concerning steam generator acoustic leak detection for the demonstration FBR plant

International Nuclear Information System (INIS)

Higuchi, Masahisa

1990-01-01

The Japan Atomic Power Co. (JAPC) started the research and development into Acoustic Leak Detection for the Demonstration FBR (D-FBR) plant in 1989. Acoustic Leak Detection is expected as a water leak detection system in the Steam Generator for the first D-FBR plant. JAPC is presently analyzing data on Acoustic Leak Detection in order to form some basic concepts and basic specifications about leak detection. Both low frequency types and high frequency types are selected as candidates for Acoustic Leak Detection. After a review of both types, either one will be selected for the D-FBT plant. A detailed Research and Development plan on Acoustic Leak Detection, which should be carried out prior to starting the construction of the D-FBR plant, is under review. (author). 3 figs, 2 tabs

Methodologies for evaluating the proliferation resistance of nuclear fuel cycles

International Nuclear Information System (INIS)

Shiotani, Hiroki; Hori, Kei-ichiro; Takeda, Hiroshi

2001-01-01

The Japan Nuclear Cycle Development Institute (JNC) believes that the development of future nuclear fuel cycle technology should be conducted with careful consideration given to non-proliferation. JNC is studying methodologies for evaluating proliferation resistance of nuclear fuel cycle technologies. However, it is difficult to establish the methodology for evaluating proliferation resistance since the results greatly depend on the assumption for the evaluation and the surrounding conditions. This study grouped factors of proliferation resistance into categories through reviewing past studies and studied the relationships between the factors. Then, this study tried to find vulnerable nuclear material (plutonium) in some FBR fuel cycles from the proliferation perspective, and calculated the time it takes to convert the materials from various nuclear fuel cycles into pure plutonium metal under some assumptions. The result showed that it would take a long time to convert the nuclear materials from the FBR fuel cycles without plutonium separation. While it is a preliminary attempt to evaluate a technical factor of proliferation resistance as the basis of the institutional proliferation resistance, the JNC hopes that it will contribute to future discussions in this area. (author)

Application of core structural design guidelines in conceptual fuel pin design

International Nuclear Information System (INIS)

Patel, M.R.; Stephen, J.D.

1979-01-01

The paper describes an application of the Draft RDT Standards F9-7, -8, and -9 to conceptual design of Fast Breeder Reactor (FBR) fuel pins. The Standards are being developed to provide guidelines for structural analysis and design of the FBR core components which have limited ductility at high fluences and are not addressed by the prevalent codes. The development is guided by a national working group sponsored by the Division of Reactor Researcch and Technology of the Department of Energy. The development program summarized in the paper includes establishment of design margins consistent with the test data and component performance requirements, and application of the design rules in various design activities. The application program insures that the quantities required for proper application of the design rules are available from the analysis methods and test data, and that the use of the same design rules in different analysis tools used at different stages of a component design producees consistent results. This is illustrated in the paper by application of the design rules in the analysis methods developed for conceptual and more detailed designs of an FBR fuel pin

Report of the evaluation by the project evaluation committee on research and development of FBR structural materials. Result intermediate evaluation in fiscal year 2002

International Nuclear Information System (INIS)

2003-03-01

The project evaluation committee of FBR (Fast Breeder Reactor) and fuel cycle evaluated the problem under the evaluation method defined by the committee on the basis of material and supplementary data produced by JNC (Japan Nuclear Cycle Development Institute) and discussion. This report consists of summary, construction of the committee, investigation process, evaluation method, opinion of object, plan, research system, results, relation to practical use, public offering of results, development and others and total evaluation. Each opinion of the committee member are reported. Development of structural material is important problem of R and D of FBR. The estimation techniques of materials, high temperature structural and earthquake-resistant design techniques are evaluated to obtain the fruitful results. The remarkable results are development of materials of 316FR, using plate to reactor, estimation method of thermal striping, improvement of earthquake-resistant analysis technology, contribution to evaluation analysis of Monju second liner and basic information to develop quantity technique of material damage. (S.Y.)

Utilization of cross-section covariance data in FBR core nuclear design and cross-section adjustment

International Nuclear Information System (INIS)

Ishikawa, Makoto

1994-01-01

In the core design of large fast breeder reactors (FBRs), it is essentially important to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of the plant. The cross-section errors, that is, covariance data are one of the most dominant sources for the prediction uncertainty of the core parameters, therefore, quantitative evaluation of covariance data is indispensable for FBR core design. The first objective of the present paper is to introduce how the cross-section covariance data are utilized in the FBR core nuclear design works. The second is to delineate the cross-section adjustment study and its application to an FBR design, because this improved design method markedly enhances the needs and importance of the cross-section covariance data. (author)

Analytical determination of thermal conductivity of W-UO2 and W-UN CERMET nuclear fuels

Science.gov (United States)

Webb, Jonathan A.; Charit, Indrajit

2012-08-01

The thermal conductivity of tungsten based CERMET fuels containing UO2 and UN fuel particles are determined as a function of particle geometry, stabilizer fraction and fuel-volume fraction, by using a combination of an analytical approach and experimental data collected from literature. Thermal conductivity is estimated using the Bruggeman-Fricke model. This study demonstrates that thermal conductivities of various CERMET fuels can be analytically predicted to values that are very close to the experimentally determined ones.

Fuel assembly

International Nuclear Information System (INIS)

Yokota, Tokunobu.

1990-01-01

A fuel assembly used in a FBR type nuclear reactor comprises a plurality of fuel rods and a moderator guide member (water rod). A moderator exit opening/closing mechanism is formed at the upper portion of the moderator guide member for opening and closing a moderator exit. In the initial fuel charging operation cycle to the reactor, the moderator exit is closed by the moderator exit opening/closing mechanism. Then, voids are accumulated at the inner upper portion of the moderator guide member to harden spectrum and a great amount of plutonium is generated and accumulated in the fuel assembly. Further, in the fuel re-charging operation cycle, the moderator guide member is used having the moderator exit opened. In this case, voids are discharged from the moderator guide member to decrease the ratio, and the plutonium accumulated in the initial charging operation cycle is burnt. In this way, the fuel economy can be improved. (I.N.)

Dynamic behaviour of FBR fuel pin bundles

International Nuclear Information System (INIS)

Martin, P.H.; Van Dorsselaere, J.P.; Ravenet, A.

1990-01-01

A programme of shock tests on a fast neutron reactor subassembly model (SPX1 geometry) including a complete bundle of fuel pins (dummy elements) is being carried out in the BELIER test facility at Cadarache. The purpose of these tests is: to determine the distribution of dynamic forces applied to the fuel rod clads under the impact conditions encountered in a reactor during a earthquake; to reduce as much as possible the conservatism of the methods presently used for the calculation of those forces. The test programme, now being completed, consists of the following steps: impacts on the mock-up in air with an non-compact bundle (situation of the subassembly at beginning of life (BOL) with clearances within the bundle); impacts under the same conditions but with fluid (water) in the subassembly; impacts on the mock-up in air and with a compacted bundle (simulating the conditions of an end-of-life (EOL) bundle with no clearance within the bundle). The accelerations studied in these tests cover the range encountered in design calculations for the subassembly frequencies in beam mode. (author)

PENENTUAN FAKTOR YANG BERPENGARUH DALAM FAULTY BEHAVIOR RISK MELALUI PENDEKATAN METODE FUZZY ANALYTIC HIERARCHY PROCESS

Directory of Open Access Journals (Sweden)

Rizal Irfan Fuadi

2015-06-01

Full Text Available Manajemen keselamatan merupakan pengorganisasian, sumber daya manusia, kebijakan dan prosedur interaktif yang bertujuan untuk mengurangi kemungkinan kerusakan dan kerugian di tempat kerja. Salah satu cara untuk memperbaiki manajemen keselamatan di perusahaan adalah dengan melakukan penelitian mengenai faktor yang berpengaruh dalam risiko kesalahan perilaku. Secara umum, faktor-faktor yang mempengaruhi keselamatan kerja tidak memiliki struktur fisik. Untuk itu, maka masalah pada kondisi nyata dapat direpresentasikan dengan cara yang lebih baik menggunakan angka fuzzy untuk mengevaluasi faktor-faktor ini. Pada penelitian ini, pendekatan Fuzzy AHP bertujuan untuk menentukan tingkat Faulty Behavior Risk (FBR pada sistem kerja.Penentuan faktor yang berpengaruh dalam Faulty Behavior Risk (FBR/risiko dari perilaku yang salah dimulai dengan menentukan responden, penyusunan kuesioner, uji validitas dan reliabilitas, hasil dari kuesioner dijadikan inputan dalam pengolahan data dengan metode FuzzyAnalytic Hierachy Process (FAHP. Kuesioner ini dibuat berdasarkan konsep safety management yang terdiri dari 4 faktor, yaitu faktor organisasi, faktor pribadi, faktor pekerjaan dan faktor lingkungan. Penelitian ini dilakukan pada Bay 2.1 yang memproduksi panel dan bay 7.1 yang memproduksi finning. Berdasarkan perhitungan FBR pada bay 2.1 menunjukkan nilai 0,4793 yang berarti risiko kesalahan perilaku di antara batas bawah (0,25 dan batas atas (0,50. Sedangkan pada bay 7.1 sebesar 0,5317 yang berarti risiko kesalahan perilaku memiliki potensi tinggi karena berada di atas batas atas. Dari hasil penentuan FBR didapatkan nilai  pada bay 2.1 yang memiliki risiko penyebab tertinggi terdapat pada sub faktor kurang persiapan (0,0788 sedangkan pada bay 7.1 dengan nilai FBR sebesar 0,5317 yang memiliki risiko tertinggi terdapat pada sub faktor kelelahan kerja (0,0970. Melalui penelitian ini, faktor penyebab kesalahan perilaku kerja dapat diketahui dan diberikan

Study of photo-stimulated luminescence in Ba0.95M0.05FBr:Eu2+ (M = Sr, Ca) powder

International Nuclear Information System (INIS)

Zhou Yingxue; Wang Dongsheng; Zhang Xinyi

2001-01-01

The luminescence centered at about 390 nm due to 4f 6 5d→4f 7 transition of Eu 2+ can be observed without X-ray or UV-light pre-irradiation for BaFBr:Eu 2+ , Ba 0.95 Sr 0.05 FBr:Eu 2+ and Ba 0.95 Ca 0.05 FBr:Eu 2+ powder samples which can be excited by light with wave length longer than 400 nm. It could be attributed to F centers which are formed in the preparation of samples. There are two broad bands in the photo-stimulated spectra. One peaked at 535 nm results from the excitation of electrons at F(F - ) center, and the other at 708 nm might be originated from the electron excitation of F(Br - ) center or F(Br - ) accumulators. If the Ba 2+ (ion's radius equals to 0.135 nm) in the samples were replaced by about 50% mol. of Sr 2+ (ion's radius equals to 0.113 nm) or Ca 2+ (ion's radius equals to 0.099 nm), the intensities of 535 nm and 708 nm bands decrease with the decreasing of ion radius. On the other hand, the shoulders at 580 nm and 650 nm or 575 nm and 645 nm can appear. In absorption spectra of 50-400 nm range, one can observe the red-shift of 79 nm, 83 nm peaks and new absorption peaks. It means that the new color centers: F(F - , Sr 2+ ), F(br - , Sr 2+ ) or F(F - , Ca 2+ ), F(br - , Ca 2+ ) are formed, when Ba 2+ were replaced with Sr 2+ or Ca 2+ in samples. It corresponds to the shoulders in photo-stimulated spectra

Seismic design for Monju FBR power plant

International Nuclear Information System (INIS)

1982-01-01

This technical report introduces the basic concept on the aseismatic design of the FBR ''Monju'' power station, of which the construction in Tsuruga is planned by the Power Reactor and Nuclear Fuel Development Corp. The safety design of Monju has been performed according to ''The concept of evaluating the safety of fast breeder reactors'', and the thought concerning the aseismatic design also is written in it. According to it, ''The guide for the examination of aseismatic design regarding power reactor facilities'' should be referred to, and the classification according to the importance in aseismatic design must be made, taking the features in the design of liquid metal-cooled FBRs fully in consideration. In the aseismatic design of Monju performed according to these basic concept, the following two points were examined. In the aseismatic design of the equipment and piping, the difference of construction from LWRs such as low pressure, thin walled and high temperature construction is taken in consideration. The classification according to the aseismatic importance of the system and equipment is made on the basis of the features in the design of Monju. The classification according to aseismatic importance, the method of calculating earthquake power, the combination of loads and the allowable limit, and the aseismatic construction of the main facilities are reported. (Kako, I.)

Study on the FBR cycle introduction scenario. 2. A study on the role of nuclear energy under the diversity of energy supply-and-demand

International Nuclear Information System (INIS)

Ohtaki, Akira; Ono, Kiyoshi; Hirao, Kazunori

2002-03-01

This report concerns it self with the results of an investigation about the possibility of future nuclear utilization in the part of FBR Cycle Introduction Scenario Study in the JNC's 'Feasibility Study on Commercialized Fast Reactor Cycle System (the F/S)'. We have investigated about the problems that confront energy industries and electric power companies, the capacities of distributed generation, the coexistence method of a distributed generation and large-scale power supply generation, and the development status of a small-scale nuclear reactor from a wide viewpoint. Especially the spread of distributed generation causes the decrease of the electricity demand which the electric power companies supplies. Since introduction scale of a distributed power supply is also expected to increase in the future, it will give some influences to a future nuclear plan and a power supply plan. The hydrogen utilization with out greenhouse gas mission is expected to spread with distributed generation, such as a fuel cell and a micro-gas turbine. Therefore, we proposed the new business model that the hydrogen produced by using nuclear surplus electricity is consumed distributed generation, such as a fuel cell and a micro-gas turbine. We plan to evaluate quantitatively the best power supply composition based on this load stability business model, FBR introduction capacities, the load factor, and the amount of CO 2 reduction. (author)

AN ANALYTICAL FRAMEWORK FOR ASSESSING RELIABLE NUCLEAR FUEL SERVICE APPROACHES: ECONOMIC AND NON-PROLIFERATION MERITS OF NUCLEAR FUEL LEASING

International Nuclear Information System (INIS)

Kreyling, Sean J.; Brothers, Alan J.; Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.

2010-01-01

The goal of international nuclear policy since the dawn of nuclear power has been the peaceful expansion of nuclear energy while controlling the spread of enrichment and reprocessing technology. Numerous initiatives undertaken in the intervening decades to develop international agreements on providing nuclear fuel supply assurances, or reliable nuclear fuel services (RNFS) attempted to control the spread of sensitive nuclear materials and technology. In order to inform the international debate and the development of government policy, PNNL has been developing an analytical framework to holistically evaluate the economics and non-proliferation merits of alternative approaches to managing the nuclear fuel cycle (i.e., cradle-to-grave). This paper provides an overview of the analytical framework and discusses preliminary results of an economic assessment of one RNFS approach: full-service nuclear fuel leasing. The specific focus of this paper is the metrics under development to systematically evaluate the non-proliferation merits of fuel-cycle management alternatives. Also discussed is the utility of an integrated assessment of the economics and non-proliferation merits of nuclear fuel leasing.

PROMILLE database as a part of JNC reactor physics analytical system for BFS-2 fast critical facility experiments analysis

International Nuclear Information System (INIS)

Bednyakov, Sergey

2000-12-01

The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS Visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-1i1 uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core. (author)

Standardization of dosimetry and damage analysis work for U.S. LWR, FBR, and MFR development program

International Nuclear Information System (INIS)

McElroy, W.N.; Doran, D.G.; Gold, R.; Morgan, W.C.; Grundl, J.A.; McGarry, E.D.; Kam, F.B.K.; Swank, J.H.; Odette, G.R.

1978-01-01

The accuracy requirements for various measured/calculated exposure and correlation parameters associated with current dosimetry and damage analysis procedures and practices depend on the accuracy needs of reactor development efforts in testing, design, safety, operations, and surveillance programs. Present state-of-the-art accuracies are estimated to be in the range of +-2 to 30 percent (1 sigma), depending on the particular parameter. There now appears to be international agreement, at least for the long term, that most reactor fuels and materials programs will not be able to accept an uncertainty greater than about +5 percent (1 sigma). The current status of dosimetry and damage analysis standardization work within the U.S. for LWR, FBR and MFR is reviewed in this paper

Optimization of FBR fuel element for high burnup

International Nuclear Information System (INIS)

Marbach, G.; Millet, P.

1985-03-01

After a brief historical background showing evolution of the French fast reactor fuel element from RAPSODIE to PHENIX and SUPER PHENIX we have examined the main points which have permitted to increase irradiation performance of the subassembly

Present condition of survey research on actualization strategy of fast breeding reactor (FBR) cycling. General outlines on the research

International Nuclear Information System (INIS)

Hara, Hideaki

2001-01-01

The Japan Nuclear Cycle Development Institute (JNC) started the survey research on actualization strategy of FBR cycling under cooperation of related organizations such as electric business company and so on, on July, 1999. The research aims at preparation of technical system to establish the FBR cycling for a future main energy supply source by extracting an actualization picture maximum activated advantages originally haven by the FBR cycling and by proposing a developmental strategy flexibly responsible to diverse needs in future society. Here was reported on effort state of its phase 1 (two years between 1999 and 2000 fiscal years). In the phase 1, it was planned to perform research and development shown as follows: 1) Extraction of actualization candidate concept on the FBR cycling under a premise of safety security and a viewpoint of evaluation on economics, resource effective usage, environmental loading reduction, and nuclear dispersion resistance by conducting investigation and evaluation of wide technical choices adopting innovative techniques, and 2) Embodiment of a research and development program of phase 2 (from 2001 to 2005 fiscal years) by investigating some technical subjects important for selection of research and development program aiming at actualization and its candidate concept on the FBR cycling. (G.K.)

Verification of the FBR fuel bundle-duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

Science.gov (United States)

Uwaba, Tomoyuki; Ito, Masahiro; Nemoto, Junichi; Ichikawa, Shoichi; Katsuyama, Kozo

2014-09-01

The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle-duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

Verification of the FBR fuel bundle–duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

Energy Technology Data Exchange (ETDEWEB)

Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Nemoto, Junichi [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ichikawa, Shoichi [Japan Atomic Energy Agency, 2-1, Shiraki, Tsuruga-shi, Fukui 919-1279 (Japan); Katsuyama, Kozo [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

2014-09-15

The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle–duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

MOX fuel use as a back-end option: Trends, main issues and impacts on fuel cycle management

International Nuclear Information System (INIS)

Fukuda, K.; Choi, J.-S.; Shani, R.; Durpel, L. van den; Bertel, E.; Sartori, E.

2000-01-01

In the past decades while the FBIULWR fuel cycle concept was zealously being developed, MOX-fuel use in thermal reactors was taken as an alternative back-end policy option. However, the plutonium recycling with LWRs has evolved to industrial level, gaining high maturity through the incubative period while FBR deployment was envisaged. Today, MOX-fuel use in LWRs makes integral part of the fuel cycle for those countries relying on the recycling policy. Developments to improve the fuel cycle performance, including the minimisation of remaining wastes, and the reactor engineering aspects owing to MOX-fuel use, are continued. This paper jointly presented by IAEA and OECD/NEA brings an integrated overview on MOX use as a back-end policy, covering MOX fuel utilisation, fuel performance and technology, economics, licensing, MOX fuel trends in the coming decades. (author)

Development of knowledge-based operator support system for steam generator water leak events in FBR plants

International Nuclear Information System (INIS)

Arikawa, Hiroshi; Ida, Toshio; Matsumoto, Hiroyuki; Kishida, Masako

1991-01-01

A knowledge engineering approach to operation support system would be useful in maintaining safe and steady operation in nuclear plants. This paper describes a knowledge-based operation support system which assists the operators during steam generator water leak events in FBR plants. We have developed a real-time expert system. The expert system adopts hierarchical knowledge representation corresponding to the 'plant abnormality model'. A technique of signal validation which uses knowledge of symptom propagation are applied to diagnosis. In order to verify the knowledge base concerning steam generator water leak events in FBR plants, a simulator is linked to the expert system. It is revealed that diagnosis based on 'plant abnormality model' and signal validation using knowledge of symptom propagation could work successfully. Also, it is suggested that the expert system could be useful in supporting FBR plants operations. (author)

Nuclear fuel assembly

International Nuclear Information System (INIS)

Ito, Arata; Wakamatsu, Mitsuo.

1976-01-01

Object: To permit the coolant in an FBR type reactor to enter from the entrance nozzle into a nuclear fuel assembly without causing cavitation. Structure: In a nuclear fuel assembly, which comprises a number of thin fuel pines bundled together at a uniform spacing and enclosed within an outer cylinder, with a handling head connected to an upper portion of the outer cylinder and an entrance nozzle connected to a lower portion of the cylinder, the inner surface of the entrance nozzle is provided with a buffer member and an orifice successively in the direction of flow of the coolant. The coolant entering from a low pressure coolant chamber into the entrance nozzle strikes the buffer member and is attenuated, and thereafter flows through an orifice into the outer cylinder. (Horiuchi, T.)

Super Phenix 1 fuel cycle, technical and economical outlooks

International Nuclear Information System (INIS)

Mougniot, J.C.; Baumier, J.; Duchatelle, L.

1982-01-01

An analysis of the costs of the various parts of the Super Phenix 1 fuel cycle is presented. The basis for calculating the mean levelized present unit cost used in French economic analyses is described. A description of the fuel cycle which follows includes the physical characteristics and management of the fuel and the costs of fuel services and raw materials. The results of calculations about Super Phenix mean levelized present fuel cycle unit cost are indicated and a comparison with two, four and six 1500 MWe units and PWR units is made. Finally conclusions are drawn about the economic possibility of FBR deployment. (U.K.)

Method of transporting fuel assemblies

International Nuclear Information System (INIS)

Okada, Katsutoshi.

1979-01-01

Purpose: To enable safety transportation of fuel assemblies for FBR type reactors by surrounding each of fuel elements in a wrapper tube by a rubbery, hollow cylindrical container and by sealing medium such as air to the inside of the container. Method: A fuel element is contained in a hollow cylindrical rubber-like tube. The fuel element has an upper end plug, a lower end plug and a wire spirally wound around the outer periphery. Upon transportation of the fuel assemblies, each of the fuel elements is covered with the container and arranged in the wrapper tube and then the fuel assemblies are assembled. Then, medium such as air is sealed for each of the fuel elements by way of an opening and then the opening is tightly closed. Before loading the transported fuel assemblies in the reactor, the medium is discharged through the opening and the container is completely extracted and removed from the inside of the wrapper tube. (Seki, T.)

Nuclear fuel reprocessing expansion strategies

International Nuclear Information System (INIS)

Gallagher, J.M.

1975-01-01

A description is given of an effort to apply the techniques of operations research and energy system modeling to the problem of determination of cost-effective strategies for capacity expansion of the domestic nuclear fuel reprocessing industry for the 1975 to 2000 time period. The research also determines cost disadvantages associated with alternative strategies that may be attractive for political, social, or ecological reasons. The sensitivity of results to changes in cost assumptions was investigated at some length. Reactor fuel types covered by the analysis include the Light Water Reactor (LWR), High-Temperature Gas-Cooled Reactor (HTGR), and the Fast Breeder Reactor (FBR)

Lining facility for FBR type reactor

International Nuclear Information System (INIS)

Shimano, Kunio.

1991-01-01

In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

Failed fuel detection device

International Nuclear Information System (INIS)

Doi, Akira.

1994-01-01

The device of the present invention concerns a failed fuel detection device for a nuclear reactor, such as an FBR type reactor, using electroconductive coolants. A sampling port is disposed at the upper portion of the fuel assembly so as to cover the assembly, so that coolants in the fuel assembly are sampled to improve a device for detecting fuel failure. That is, when coolants in the fuel assembly are sampled from the sampling port, the flow of electroconductive coolants in an sampling tube is detected by a flowmeter, to control an electromagnetic pump. The flow of electroconductive coolants is stopped against the waterhead pressure and dynamic pressure of the conductive coolants, and a predetermined amount of the coolants is pumped up to the sampling tank. Gas is supplied to the pumped up coolants so that fissile products are transferred from the coolants to a gas phase. Radiation in the gas in a gas recycling system is measured to detect presence of fuel failure. (I.S.)

Principles of MONJU maintenance. Characteristic of MONJU maintenance and reflection of LWR maintenance experience to FBR

International Nuclear Information System (INIS)

Nakai, Satoru; Nishio, Ryuichi; Uchihashi, Masaya; Kaneko, Yoshihisa; Yamashita, Hironobu; Yamaguchi, Atsunori; Aoki, Takayuki

2014-01-01

A sodium cooled fast breeder reactor (FBR) has unique systems and components and different degradation mechanism from light water reactor (LWR) so that need to establish maintenance technology in accordance with its features. The examination of the FBR maintenance technology is carried out in the special committee for considering the maintenance for Monju established in the Japan Society of Maintenology (JSM). As a result of the study such as extraction of Monju maintenance feature, maintenance technology benchmark between Monju and LWR components and survey of LWR maintenance experience, it is clear that principles of maintenance are same as LWR, necessity of LWR maintenance experience reflection and points to be considered in Monju maintenance. The road map to establish a FBR maintenance technology in the technical aspect became clear and it is vital to acquire operation and maintenance experience of the plant to implement this road map, and to establish a fast reactor maintenance. (author)

Analytical evaluation of local fault in sodium cooled small fast reactor (4S). Preliminary evaluation of partial blockage in coolant channel

International Nuclear Information System (INIS)

Nishimura, Satoshi; Ueda, Nobuyuki

2007-01-01

Local faults are fuel failures that result from heat removal imbalance within a single subassembly especially in FBRs. Although the occurrence frequency of local faults is quite low, the licensing body required local faults evaluations in previous FBR plants to confirm the potential for the occurrence of severe fuel subassembly failure and its propagation. A conceptual design of 4S (Super-Safe, Small and Simple) is a sodium cooled fast reactor, which aims at an application to dispersed energy source and long core lifetime. It has a dense arrangement of fuel pins to achieve a long lifetime. Therefore, from the viewpoint of thermal hydraulics, the 4S reactor is considered to have more potential for coolant boiling and fuel pin failure caused by formation of local blockage, comparing these potential in the conventional FBRs. The objective of the present study is to evaluate the effect of local blockage on the coolant flow pattern and temperature rise in the 4S-type fuel subassembly under the normal operation condition. A series of three-dimensional thermal-hydraulic analysis in a single subassembly with local blockage was conducted by the commercialized CFD code 'PHOENICS'. Analytical results show that the peak coolant temperature behind the blockage rises with increasing the blockage area, however, the coolant boiling does not occur under the present analytical conditions. On the other hand, it is found that the liquid phase formation caused by eutectic reactions will occur between the metallic fuel and the cladding under the local blockage condition. However, the penetration rate of liquid phase at fuel-cladding interface is quit low. Therefore, it is expected that rapid fuel pin failure and its propagation to surrounding pins due to liquid phase formation will not occur. (author)

Investigation of the fuel temperature evaluation method at BOL

International Nuclear Information System (INIS)

Ishii, Tetsuya; Asaga, Takeo; Nemoto, Junichi

1999-06-01

It is one of the major subjects in the improvement of the design method for determining the thermal conditions of the solid type Mixed - Oxide (MOX) fuels in FBR to evaluate the fuel temperature at BOL as precisely as possible. Therefore, we have planned to modify the fuel temperature evaluation method 'FEVER', which was developed by JNC in 1988, as one of the investigation for the establishment of the precise fuel temperature evaluation method. And, we also have planned to use the modified FEVER, named FEVER-M', for estimation of the irradiation conditions of the PTM test in Joyo, called 'B10 test', planning to perform in 2000. In this work, the following results were obtained; 1) As a result of the modification, the uncertainty in the fuel temperature evaluation of 'FEVER-M' is reduced to about ±60 K. 2) Estimating the irradiation conditions of 'B10' test using the method 'FEVER-M', it is found that the appropriate maximum linear heat rate for the test is 620 W/cm. The detail plans of the 'B10' test were also determined based on the results. 3) Based on the results of this work, it is found that one of the effective procedure for the improvement of the accuracy of the fuel temperature evaluation method seems to calculate the fuel temperature taking the pellet relocation phenomena into account. In future, although there are a lot of matters to be discussed in this phenomena, the design method for the thermal conditions of the MOX fuels in FBR should be performed with taking the pellet relocation phenomena into account. (author)

FBR type reactors

International Nuclear Information System (INIS)

Suzuoki, Akira; Yamakawa, Masanori.

1985-01-01

Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

Analytical criteria for fuel failure modes observed in reactivity initiated accidents

International Nuclear Information System (INIS)

Luxat, J.C.

2005-01-01

The behaviour of nuclear fuel subjected to a short duration power pulse is of relevance to LWR and CANDU reactor safety. A Reactivity Initiated Accident (RIA) in an LWR would subject fuel to a short duration power pulse of large amplitude, whereas in CANDU a large break Loss of Coolant Accident (LOCA) would subject fuel to a longer duration, lower amplitude power excursion. The energy generated in the fuel during the power pulse is a key parameter governing the fuel response. This paper reviews the various power pulse tests that have been conducted in research reactors over the past three decades and summarizes the fuel failure modes that that have been observed in these tests. A simple analytical model is developed to characterize fuel behaviour under power pulse conditions and the model is applied to assess the experimental data from the power pulse tests. It is shown that the simple model provides a good basis for establishing criteria that demarcate the observed fuel failure modes for the various fuel designs that have been used in these tests. (author)

Simulation of Fungal-Mediated Cell Death by Fumonisin B1 and Selection of Fumonisin B1–Resistant (fbr) Arabidopsis Mutants

Science.gov (United States)

Stone, Julie M.; Heard, Jacqueline E.; Asai, Tsuneaki; Ausubel, Frederick M.

2000-01-01

Fumonisin B1 (FB1), a programmed cell death–eliciting toxin produced by the necrotrophic fungal plant pathogen Fusarium moniliforme, was used to simulate pathogen infection in Arabidopsis. Plants infiltrated with 10 μM FB1 and seedlings transferred to agar media containing 1 μM FB1 develop lesions reminiscent of the hypersensitive response, including generation of reactive oxygen intermediates, deposition of phenolic compounds and callose, accumulation of phytoalexin, and expression of pathogenesis-related (PR) genes. Arabidopsis FB1-resistant (fbr) mutants were selected directly by sowing seeds on agar containing 1 μM FB1, on which wild-type seedlings fail to develop. Two mutants chosen for further analyses, fbr1 and fbr2, had altered PR gene expression in response to FB1. fbr1 and fbr2 do not exhibit differential resistance to the avirulent bacterial pathogen Pseudomonas syringae pv maculicola (ES4326) expressing the avirulence gene avrRpt2 but do display enhanced resistance to a virulent isogenic strain that lacks the avirulence gene. Our results demonstrate the utility of FB1 for high-throughput isolation of Arabidopsis defense-related mutants and suggest that pathogen-elicited programmed cell death of host cells may be an important feature of compatible plant–pathogen interactions. PMID:11041878

Thorium fuel cycle analysis

Energy Technology Data Exchange (ETDEWEB)

Yamaji, K [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

1980-07-01

Systems analysis of the thorium cycle, a nuclear fuel cycle accomplished by using thorium, is reported in this paper. Following a brief review on the history of the thorium cycle development, analysis is made on the three functions of the thorium cycle; (1) auxiliary system of U-Pu cycle to save uranium consumption, (2) thermal breeder system to exert full capacity of the thorium resource, (3) symbiotic system to utilize special features of /sup 233/U and neutron sources. The effects of the thorium loading in LWR (Light Water Reactor), HWR (Heavy Water Reactor) and HTGR (High Temperature Gas-cooled Reactor) are considered for the function of auxiliary system of U-Pu cycle. Analysis is made to find how much uranium is saved by /sup 233/U recycling and how the decrease in Pu production influences the introduction of FBR (Fast Breeder Reactor). Study on thermal breeder system is carried out in the case of MSBR (Molten Salt Breeder Reactor). Under a certain amount of fissile material supply, the potential system expansion rate of MSBR, which is determined by fissile material balance, is superior to that of FBR because of the smaller specific fissile inventory of MSBR. For symbiotic system, three cases are treated; i) nuclear heat supply system using HTGR, ii) denatured fuel supply system for nonproliferation purpose, and iii) hybrid system utilizing neutron sources other than fission reactor.

Analytical Dancoff factor evaluations for reactor designs loaded with TRISO particle fuel

International Nuclear Information System (INIS)

Ji, Wei; Liang, Chao; Pusateri, Elise N.

2014-01-01

Highlights: • The Dancoff factors for randomly distributed TRISO fuel particles are evaluated. • A new “dual-sphere” model is proposed to predict Dancoff factors. • The new model accurately accounts for the coating regions of fuel particles. • High accuracy is achieved over a broad range of design parameters. • The new model can be used to analyze reactors with double heterogeneity. - Abstract: A new mathematical model, the dual-sphere model, is proposed to analytically evaluate Dancoff factors of TRISO fuel kernels based on the chord method. The accurate evaluation of fuel kernel Dancoff factors is needed when one analyzes nuclear reactors loaded with TRISO particle fuel. In these reactor designs, fuel kernels are randomly distributed and shield each other, causing a shadowing effect. The Dancoff factor is a quantitative measure of this effect and is determined by the spatial distribution of fuel kernels. A TRISO fuel particle usually consists of four layers that form a coating region outside the fuel kernel. When fuel particles are loaded in the reactor, the spatial distribution of fuel kernels can be affected by the thickness of the coating region. Therefore, the coating region should be taken into account in the calculation of Dancoff factors. However, the previous model, the single-sphere model, assumes no coating regions in the Dancoff factor predictions. To address this model deficiency, the dual-sphere model is proposed by deriving a new chord length distribution function between two fuel kernels that explicitly accounts for coating regions. The new model is employed to derive analytical solutions of infinite medium, intra-fuel pebble and intra-fuel compact/pin Dancoff factors over a wide range of volume packing fractions of TRISO fuel particles, varying from 2% to 60%. Comparisons are made with the predictions from the single-sphere model and reference Monte Carlo simulations. A significant improvement of the accuracy, over the ranges of

Investigations of chemical reactions between U-Zr alloy and FBR cladding materials

International Nuclear Information System (INIS)

Ishii, Tetsuya; Ukai, Shigeharu

2005-07-01

U-Pu-Zr alloys are candidate materials for commercial FBR fuel. However, informations about chemical reactions with cladding materials developed by JNC for commercial FBR have not been well obtained. In this work, the reaction zones formed in four diffusion couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS, U-10wt.%Zr/12Cr-ODS, and U-10wt.%Zr/Fe at about 1013K have been examined and following results were obtained. 1) At about 1013K, in the U-10wt.%Zr/Fe couple, the liquid phase zones were obtained. In the other couples U-10wt.%Zr/PNC-FMS, U-10wt.%Zr/9Cr-ODS and U-10wt.%Zr/12Cr-ODS, no liquid phase zones were obtained. The obtained chemical reaction zones in the later 3 couples were similar to the reported ones obtained in U-Zr/Fe couples without liquid phase formation. In comparison with the reaction zones obtained in the U-10wt.%Zr/Fe couple, the reaction zones inside cladding materials obtained in the PNC-FMS, 9Cr-ODS, and 12Cr-ODS couples were thin. 2) From the investigations of relationship between the obtained depths of the chemical reaction zones inside cladding materials and composition of the cladding materials, it was considered that the depth of chemical reaction zone would depend on the Cr content of the cladding materials and the depth would decrease with increasing Cr content, resulting in prevention of liquid phase formation. 3) From the investigations of the mechanisms of chemical reactions between U-Pu-Zr/cladding materials, it was considered that the same effect of Cr obtained in the U-Zr/cladding materials would be expected in U-Pu-Zr/cladding materials. Those seemed to indicate that the threshold temperatures of liquid phase formation for U-Pu-Zr/PNC-FMS, U-Pu-Zr/9Cr-ODS, and U-Pu-Zr/12Cr-ODS might be higher than that for U-Pu-Zr/Fe. (author)

Manufacturing method for fuel assembly

International Nuclear Information System (INIS)

Yamaguchi, Takashi.

1997-01-01

In an FBR type reactor, uranium/plutonium mixed oxide fuels (MOX fuels) are used. Nuclear fuel materials containing uranium and plutonium are filled to a portion or all of a plurality of fuel rods. In this case, an equivalent fissile coefficient (B) based on a combustion guarantee method defined by the formula: (B) = (M) · (F) is determined. (M) is a combustion matrix constituted based on the solution of equation of combustion which is a differential equation representing change with time of each of nuclear fuel materials during combustion. (F) is an equivalent fissile coefficient based on a reactivity keeping method which is a coefficient representing a reactivity worth equivalent with plutonium-239. The content of each of the nuclear fuel materials is determined so that the effective multiplication factor at the final stage of the operation cycle is substantially constant by using the equivalent fissile coefficient (B) based on the combustion guarantee method. (I.N.)

Evaluation of thermal physical properties for fast reactor fuels. Melting point and thermal conductivities

International Nuclear Information System (INIS)

Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki; Ogasawara, Masahiro; Tamura, Tetsuya; Sugata, Hirotada; Sunaoshi, Takeo; Shibata, Kazuya

2006-10-01

Japan Atomic Energy Agency has developed a fast breeder reactor (FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio (O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Neumann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content. (author)

Design study on advanced nuclear fuel recycling system by pyrometallurgical reprocessing technology

Energy Technology Data Exchange (ETDEWEB)

Kasai, Yoshimitsu; Kakehi, Isao; Moro, Satoshi; Tobe, Kenji; Kawamura, Fumio; Higashi, Tatsuhiro; Yonezawa, Shigeaki [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Yoshiuji, Takahiro

1998-12-01

The Japan Nuclear Fuel Cycle Development Institute is conducting research and development on the nuclear fuel recycling system, which will improve the economy, safety, and environmental impact of the nuclear fuel recycling system in the age of the FBR. The System Engineering Division in the O-arai Engineering Center has conducted a design study on an advanced nuclear fuel recycling system for FBRs by using pyrometallurgical reprocessing technology. The system is an economical and compact module-type system, and can be used for reprocessing oxide fuel and also new types of fuel (metal fuel and nitride fuel). This report describes the concept of this system and results of the design study. (author)

Design study on advanced nuclear fuel recycling system by pyrometallurgical reprocessing technology

International Nuclear Information System (INIS)

Kasai, Yoshimitsu; Kakehi, Isao; Moro, Satoshi; Tobe, Kenji; Kawamura, Fumio; Higashi, Tatsuhiro; Yonezawa, Shigeaki; Yoshiuji, Takahiro

1998-01-01

The Japan Nuclear Fuel Cycle Development Institute is conducting research and development on the nuclear fuel recycling system, which will improve the economy, safety, and environmental impact of the nuclear fuel recycling system in the age of the FBR. The System Engineering Division in the O-arai Engineering Center has conducted a design study on an advanced nuclear fuel recycling system for FBRs by using pyrometallurgical reprocessing technology. The system is an economical and compact module-type system, and can be used for reprocessing oxide fuel and also new types of fuel (metal fuel and nitride fuel). This report describes the concept of this system and results of the design study. (author)

Development of new decladding system in the reprocessing process for FBR fuel

International Nuclear Information System (INIS)

Yamada, Seiya; Washiya, Tadahiro; Takeuchi, Masayuki; Koizumi, Tsutomu; Aose, Shinichi

2005-01-01

As a part of the feasibility study on commercialized fast reactor cycle systems, Japan Nuclear Cycle Development Institute (JNC) has been developing the fuel decladding technology for the dry reprocessing process (oxide electrowinning process) and aqueous reprocessing process. Particularly, in the oxide electrowinning process, the spent fuel should be reduced to powder for quick dissolution in the molten salt at electrolyzer. Therefore, JNC proposes new decladding system with innovative mechanical decladding devices. The decladding system consists of fuel crushing stage, hull separation stage and hull rinsing stage. In the fuel crushing stage, disassembled spent fuel pins are crushed and powdered by mechanical decladding device, then the following stage, the hull and the fuel powder are separated by magnetic separator. Only the fuel powder is fed to the electrolyzer. On the other side, the separated hull is melted by induction heating method, and the small amount of oxide included in the hull fragments is recovered at the hull rinsing stage. The recovered oxide fuel is fed back to the electrolyzer. In this paper, the basic performance of the element equipment that composes this new decladding system will be described. (author)

Investigation on fuel-cladding chemical interaction in metal fuel for FBR

International Nuclear Information System (INIS)

Inagaki, Kenta; Nakamura, Kinya; Ogata, Takanari; Uwaba, Tomoyuki

2013-01-01

During steady-state irradiation of metallic fuel in fast reactors, rare-earth fission products can react with stainless steel cladding at the fuel-cladding interface. The authors conducted isothermal annealing tests with some diffusion couples to investigate the structure of the wastage layer formed at the interface. Candidate cladding alloys, ferritic-martensitic steel (PNC-FMS) and oxide-dispersion-strengthened (ODS) steel were assembled with rare-earth alloys, RE5 : La-Ce-Pr-Nd-Sm, which simulate the fission yield of rare-earth fission products. The diffusion couples were isothermally annealed in the temperature range of 500-650°C for up to 170 h. In both RE5/ODS-steel and RE5/PNC-FMS couples, the wastage layer of the two-phase region of the (Fe, Cr) 17 RE 2 matrix phase with the precipitation of the (Fe, RE, Cr) phase was formed. The structure was similar to that formed in RE5/Fe-12Cr and RE5/HT9 couples, which implies that the reaction between REs and steel is not significantly influenced by the minor alloying elements within the candidate cladding materials. It was also clarified that the increase in the wastage layer thickness was diffusion-controlled. The temperature dependence of the reaction rate constants were formulated, which can be the basis for the quantification of the wastage layer growth. (author)

Reactor structure for FBR

International Nuclear Information System (INIS)

Yamazaki, Tsukasa.

1986-01-01

Purpose: To prevent deformation of equipment for FBR structure by inhibiting free convection generated at the roof slab through device. Constitution: The labyrinth is placed between the lower part of the roof slab and the lower one of the positioning flange, and then, convection-preventive wrinkle is provided for the side wall for the positioning flange against the roof slab side wall. The upper part of the positioning flange is fixed to the upper surface of the roof slab, the through-device flange is connected to the lower flange, and prevent generation of thermal stress. Thus, free convection at the through-device is prevented, and it has become possible to miniaturize the seal section of the intermediate heat exchanger and prevent galling of the circulating pump. The joint position of the positioning flange with the through-device flange can be shifted to the same height level of the roof slab, and the length under the hook of the overhead crance can be reduced. (Horiuchi, T.)

Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1989-01-01

Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

Interrelationship betwen material strength and component design under elevated temperature for FBR

International Nuclear Information System (INIS)

Nakagawa, Y.

Structural design under elevated temperature for fast breeder reactor plant is very troublesome compared to that of for lower temperature. This difficulty can be mainly discussed from two different stand points. One is design and design code, another is material strength. Components in FBR are operated under creep regime and time dependent creep behaviour should be elevated properly. This means the number and combinations of design code and material strength are significantly large and makes these systems very complicated. Material selection is, in no words, not an easy job. This should be done by not only material development but also component design stand point. With valuable experience of construction and research on FBR, a lot of information on component design and material behaviour is available. And it is a time to choose the ''best material'' from the entire stand points of component construction. (author)

Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

Energy Technology Data Exchange (ETDEWEB)

Sugino, Kazuteru [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwai, Takehiko

1998-07-01

A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

FBR metallic materials test manual (English version)

International Nuclear Information System (INIS)

Odaka, Susumu; Kato, Shoichi; Yoshida, Eiichi

2003-06-01

For the development of the fast breeder reactor, this manual describes the method of in-air and in-sodium material tests and the method of organization the data. This previous manual has revised in accordance with the revision of Japanese Industrial Standard (JIS) and the conversion to the international unit. The test methods of domestic committees such as the VAMAS (Versailles Project on Advanced Materials and Standards) workshop were also refereed. The material test technologies accumulated in this group until now were also incorporated. This English version was prepared in order to provide more engineers with the FBR metallic materials test manual. (author)

FBR type reactors

International Nuclear Information System (INIS)

Nakamura, Tsugio.

1986-01-01

Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

Electronic manual of the nuclear characteristics analysis code-set for FBR

International Nuclear Information System (INIS)

Makino, Tohru

2001-03-01

Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

Investigation into rationalization of low decontamination pellet fuel fabrication plant configuration

International Nuclear Information System (INIS)

Maekawa, Kazuhiko; Yoshimura, Tadahiro; Hoshino, Yasushi; Munekata, Hideki; Tamaki, Yoshihisa

2005-02-01

In feasibility studies on commercialized FBR cycle system, a comprehensive system investigation and properties evaluation for candidate FBR cycle systems has been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation into rationalization of low decontamination pellet fuel fabrication plant configuration was carried out. Until last fiscal year, conceptual design studies of the fuel fabrication plant in 200t-HM/y scale were conducted, and system properties data concerning economics and environmental burden reduction of fuel fabrication plant was acquired. In addition to this, 50t-HM/y scale plant was also schematically studied. In this fiscal year, a rationalization study on conceptual design of 50t-HM/y scale plant was conducted with main aim of economic improvement, and the 200t-HM/y scale plant design was revised based on the recent R and D progress. The system properties data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In both case of the 50t-HM/y and 200t-HM/y scale plant, it was suggested that the equipment costs were reduced in several percentages because of reduction of maintenance equipments and cut in line number at the pellet fabrication process although granulation process fro denitration converted powder and O/M control process for pellets were added. System properties data for comparative evaluation of candidate fuel fabrication systems was also prepared. (author)

Spent fuel management in Japan

International Nuclear Information System (INIS)

Shirahashi, K.; Maeda, M.; Nakai, T.

1996-01-01

Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

Fabrication of uranium-plutonium mixed nitride fuel pins (88F-5A) for first irradiation test at JMTR

International Nuclear Information System (INIS)

Suzuki, Yasufumi; Iwai, Takashi; Arai, Yasuo; Sasayama, Tatsuo; Shiozawa, Ken-ichi; Ohmichi, Toshihiko; Handa, Muneo

1990-07-01

A couple of uranium-plutonium mixed nitride fuel pins was fabricated for the first irradiation tests at JMTR for the purpose of understanding the irradiation behavior and establishing the feasibility of nitride fuels as advanced FBR fuels. The one of the pins was fitted with thermocouples in order to observe the central fuel temperature. In this report, the fabrication procedure of the pins such as pin design, fuel pellet fabrication and characterizations, welding of fuel pins, and inspection of pins are described, together with the outline of the new TIG welder installed recently. (author)

Report of 5th new nuclear fuel research meeting, Yayoi Research Group. Trend of advanced basic research in nuclear fuel technical development

International Nuclear Information System (INIS)

1994-03-01

Theme of this meeting is 'Trend of advanced basic research in nuclear fuel technical development', and it was attempted to balance both sides of the basic research and the development. At the meeting, lectures were given on the chemical form of FPs in oxide fuel pins, the absorption of hydrogen of fuel cladding tubes, the application of hydride fuel to thorium cycle, the thermal properties of fuel cladding tubes, the preparation of NpN and heat conductivity, the high temperature chemical reprocessing of nitride fuel, the research on the annihilation treatment of minor actinide in fast reactors, the separation of TRU by dry process and the annihilation using a metallic fuel FBR. In this report, the summaries of the lectures are collected, and also the program of the meeting and the list of attendants are shown. (K.I.)

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

Energy Technology Data Exchange (ETDEWEB)

Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

1994-12-31

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

International Nuclear Information System (INIS)

Boussard, F.; Huillery, R.

1994-01-01

This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

A study on the design method of the laminated rubber bearing for FBR

International Nuclear Information System (INIS)

Mazda, T.; Ishida, K.; Ikeuchi, K.; Yashiro, T.

1994-01-01

The Central Research Institute of Electric Power Industry (Japan) has been carrying out the Demonstration Test and Research Program of the Seismic Isolation System for the Fast Breeder Reactor (FBR) under contract with the Ministry of International Trade and Industry (FY1987--FY1993). In this research program, development of the seismic isolation element, reliability evaluation of the seismic isolation system, and examination of a design-based earthquake have been conducted in order to develop and propose design guidelines of seismic isolation system for FBR plant. The purpose of this paper is to describe the test results concerned with structural design of the laminated rubber bearing, and to explain the construction of the design method for the laminated rubber bearing, which is one of the most important factors in the design guidelines of the seismic isolation system

The French R and D programme for fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Auchapt, P.; Bourgeois, M.; Calame-Longjean, A.; Miquel, P.; Sauteron, J.

1979-01-01

The process employed is the Purex process adapted to the specific case of fast breeder reactor fuels. The results achieved have demonstrated that the aqueous method can be applied to these fuels: nearly ten years of operation in the ATl workshop which reprocesses RAPSODIE fuels, and the good results obtained at the Marcoule pilot facility on large batches of fuel attest to this achievement. The CEA effort continues principally on extrapolation to industrial scale, thanks mainly to experiments conducted on industrial prototypes and to the launching of the TOR project, which will, as of 1984, allow reprocessing of FBR fuels on a significant scale, and which will provide extensive additional resources for R and D activities

Neutronics optimization of LiPb-He dual-cooled fuel breeding blanket for the fusion-driven sub-critical system

International Nuclear Information System (INIS)

Zheng Shanliang; Wu Yican

2002-01-01

The concept of the liquid Li 17 Pb 83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239 Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

Design study and evaluation of fuel fabrication systems for FR fuel cycle

International Nuclear Information System (INIS)

Namekawa, Takashi; Tanaka, Kenya; Kawaguchi, Koichi; Koike, Kazuhiro; Shimuta, Hiroshi; Suzuki, Yoshihiro

2004-01-01

The plant concept for each FBR fuel fabrication system has been constructed and evaluated, which achieves economical improvement, decrease in the environmental burden, better resource utilization, and proliferation resistance by the various innovative techniques employed. The results are as follows: (1) For oxide fuels, the simplified pelletizing method has a high technical feasibility, and it is possible to apply this method to practical process at early stage, because this method is based on wealth results of a conventional method. (2) For oxide fuels, the sphere packing fuel fabrication system by gelation and vibro-compaction processes has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the process. However this system shoulders additional cost for the liquid waste treatment process to dispose a large bulk of process liquid waste. (3) For the metal fuel, the casting system is generally expected to have high economical efficiency even for small-scale facilities, although verification for fabrication of the TRU alloy slug is required. (author)

Development of an analytical model to assess fuel property effects on combustor performance

Science.gov (United States)

Sutton, R. D.; Troth, D. L.; Miles, G. A.; Riddlebaugh, S. M.

1987-01-01

A generalized first-order computer model has been developed in order to analytically evaluate the potential effect of alternative fuels' effects on gas turbine combustors. The model assesses the size, configuration, combustion reliability, and durability of the combustors required to meet performance and emission standards while operating on a broad range of fuels. Predictions predicated on combustor flow-field determinations by the model indicate that fuel chemistry, as defined by hydrogen content, exerts a significant influence on flame retardation, liner wall temperature, and smoke emission.

Measurement of burnup in FBR MOX fuel irradiated to high burnup

International Nuclear Information System (INIS)

Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

2003-01-01

The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

Nuclear Materials Characterization in the Materials and Fuels Complex Analytical Hot Cells

International Nuclear Information System (INIS)

Rodriquez, Michael

2009-01-01

As energy prices skyrocket and interest in alternative, clean energy sources builds, interest in nuclear energy has increased. This increased interest in nuclear energy has been termed the 'Nuclear Renaissance'. The performance of nuclear fuels, fuels and reactor materials and waste products are becoming a more important issue as the potential for designing new nuclear reactors is more immediate. The Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Analytical Laboratory Hot Cells (ALHC) are rising to the challenge of characterizing new reactor materials, byproducts and performance. The ALHC is a facility located near Idaho Falls, Idaho at the INL Site. It was built in 1958 as part of the former Argonne National Laboratory West Complex to support the operation of the second Experimental Breeder Reactor (EBR-II). It is part of a larger analytical laboratory structure that includes wet chemistry, instrumentation and radiochemistry laboratories. The purpose of the ALHC is to perform analytical chemistry work on highly radioactive materials. The primary work in the ALHC has traditionally been dissolution of nuclear materials so that less radioactive subsamples (aliquots) could be transferred to other sections of the laboratory for analysis. Over the last 50 years though, the capabilities within the ALHC have also become independent of other laboratory sections in a number of ways. While dissolution, digestion and subdividing samples are still a vitally important role, the ALHC has stand alone capabilities in the area of immersion density, gamma scanning and combustion gas analysis. Recent use of the ALHC for immersion density shows that extremely fine and delicate operations can be performed with the master-slave manipulators by qualified operators. Twenty milligram samples were tested for immersion density to determine the expansion of uranium dioxide after irradiation in a nuclear reactor. The data collected confirmed modeling analysis with very tight

Study on coated layer material performance of coated particle fuel FBR (2). High temperature property and capability of coating to thick layer of TiN

International Nuclear Information System (INIS)

Naganuma, Masayuki; Mizuno, Tomoyasu

2002-08-01

'Helium Gas Cooled Coated Particle Fuel FBR' is one of attractive core concepts in the Feasibility Study on Commercialized Fast Reactor Cycle System in Japan, and the design study is presently proceeded. As one of key technologies of this concept, the coated layer material is important, and ceramics is considered to be a candidate material because of the superior refractory. Based on existing knowledge, TiN is regarded to be a possible candidate material, to which some property tests and evaluations have been conducted. In this study, preliminary tests about the high temperature property and the capability of thick layer coating of TiN have been conducted. Results of these tests come to the following conclusions. Heating tests of two kinds of TiN layer specimens coated by PVD (Physical Vapor Deposition) and CVD (Chemical Vapor Deposition) were conducted. As a result, as for CVD coating specimens, remarkable charge was not observed on the layer up to 2,000degC, therefore we concluded that the layer by CVD had applicability up to high temperature of actual operation level. On the other hand, as for PVD coating specimens, an unstable behavior that the layer changed to a mesh like texture was observed on a 2,000degC heated specimen, therefore the applied PVD method is not considered to be promising as the coating technique. The surface conditions of some parts inside CVD device were investigated in order to evaluate possibility of TiN thick coating (∼100 μm). As a result, around 500 μm of TiN coating layer was observed on the condition of multilayer. Therefore, we conclude that CVD has capability of coating up to thick layer in actual coated particle fuel fabrication. (author)

An investigation on technical feasibilities of fuel cycle for high temperature gas-cooled reactor (Case study)

International Nuclear Information System (INIS)

Sumita, Junya; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sawa, Kazuhiro

2008-03-01

In accordance with the basic policy of effectively using nuclear fuel resources, the FBR cycle, one of the most possible fuel cycle in the future, will be adapted after plu-thermal program by LWR in Japanese nuclear cycle plan. In this paper, a case study of technical investigation of HTGR fuel cycle based on HTGR fuel cycle proposed to adapt to Japanese nuclear fuel cycle plan were carried out from the viewpoint of effective utilization of uranium, fabrication technologies of MOX fuel, reprocessing technologies, amount of interim storage of HTGR fuel and graphite waste. As a result, the fuel cycle for HTGR is expected to be possible technically. (author)

Structural dynamics in FBR

International Nuclear Information System (INIS)

Bhoje, S.B.

2003-01-01

In view of thin walled large diameter shell structures with associated fluid effects, structural dynamics problems are very critical in a fast breeder reactor. Structural characteristics and consequent structural dynamics problems in typical pool type Fast Breeder Reactor are highlighted. A few important structural dynamics problems are pump induced as well as flow induced vibrations, seismic excitations, pressure transients in the intermediate heat exchangers and pipings due to a large sodium water reaction in the steam generator, and core disruptive accident loadings. The vibration problems which call for identification of excitation forces, formulation of special governing equations and detailed analysis with fluid structure interaction and sloshing effects, particularly for the components such as PSP, inner vessel, CP, CSRDM and TB are elaborated. Seismic design issues are presented in a comprehensive way. Other transient loadings which are specific to FBR, resulting from sodium-water reaction and core disruptive accident are highlighted. A few important results of theoretical as well as experimental works carried out for 500 MWe Prototype Fast Breeder Reactor (PFBR), in the domain of structural dynamics are presented. (author)

Advanced analysis technology for MOX fuel

International Nuclear Information System (INIS)

Hiyama, T.; Kamimura, K.

1997-01-01

PNC has developed MOX fuels for advanced thermal reactor (ATR) and fast breeder reactor (FBR). The MOX samples have been chemically analysed to characterize the MOX fuel for JOYO, MONJU, FUGEN and so on. The analysis of the MOX samples in glove box has required complicated and highly skilled operations. Therefore, for quality control analysis of the MOX fuel in a fabrication plant, simple, rapid and accurate analysis methods are necessary. To solve the above problems instrumental analysis and techniques were developed. This paper describes some of the recent developments in PNC. 2. Outline of recently developed analysis methods by PNC. 2.1 Determination of oxygen to metal atomic ratio (O/M) in MOX by non-dispersive infrared spectrophotometry after inert gas fusion. 7 refs, 9 figs, 4 tabs

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

Some difference of concepts between design guideline for FBR base isolation system and aseismic design guideline of LWR in Japan

International Nuclear Information System (INIS)

Shibata, Heki

1992-01-01

This paper deals with the concept and the relation of 'the Base Isolation System and FBR' to the Safety Criteria and the Guideline of the Aseismic Design of LWR in Japan. The Central Research Institute of Electric Power Industries have been working for FBR last several years. The author has been contribute to their works, and this is one of the subjects. He described his own idea obtained through the cooperative work with CRIEPI. (author)

Preliminary Analysis of the Bundle-Duct Interaction for the fuel of SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-10-15

BDI (Bundle-Duct Interaction) occurs in the fuel of SFR (Sodium-cooled Fast Reactor) due to the radial expansion and bowing of a fuel pin bundle. Under the BDI condition, excess cladding strain and hot spots would occur. Therefore, BDI, which is the dominant deformation mechanisms in a fuel pin bundle, should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE and BMBOO, have been developed to evaluate the BDI behavior. The bundle duct interaction model is also being developed for SFR in Korea. This model is based on ANSYS. In this paper, the fuel pin configuration model for the BDI calculation was established. The preliminary analysis of the bundle-duct interaction was performed to evaluate the fuel design concept.

Evaluation of core distortion in FBR

International Nuclear Information System (INIS)

Ikarimoto, I.; Tanaka, M.; Okubo, Y.

1984-01-01

The analyses of FBR's core distortion are mainly performed in order to evaluate the following items: 1) Change of reactivity; 2) Force at pads on core assemblies; 3) Withdrawal force at refueling; 4) Loading, refueling and residual deviations of wrapper tubes (core assemblies) at the top; 5) Bowing modes of guide tubes for control rods. The analysis of core distortion are performed by using computer program for two-dimensional row deformation analysis or three-dimensional core deformation if necessary, considering these evaluated items which become design conditions. This report shows the relationship between core deformation analysis and component design, a point of view of choosing an analysis program for design considering core characteristics, and computing examples of core deformation of prototype class reactor by the above code. (author)

Revisit of analytical methods for the process and plant control analyses during reprocessing of fast reactor fuels

International Nuclear Information System (INIS)

Subba Rao, R.V.

2016-01-01

CORAL (COmpact facility for Reprocessing of Advanced fuels in Lead cell) is an experimental facility for demonstrating the reprocessing of irradiated fast reactor fuels discharged from the Fast Breeder Test Reactor (FBTR). The objective of the reprocessing plant is to achieve nuclear grade plutonium and uranium oxides with minimum process waste volumes. The process flow sheet for the reprocessing of spent Fast Reactor Fuel consists of Transport of spent fuel, Chopping, Dissolution, Feed conditioning, Solvent Extraction cycle, Partitioning Cycle and Re-conversion of Plutonium nitrate and uranium nitrate to respective oxides. The efficiency and performance of the plant to achieve desired objective depends on the analyses of various species in the different steps adopted during reprocessing of fuels. The analytical requirements in the plant can be broadly classified as 1. Process control Analyses (Analyses which effect the performance of the plant- PCA); 2. Plant control Analyses (Analyses which indicates efficiency of the plant-PLCA); 3. Nuclear Material Accounting samples (Analyses which has bearing on nuclear material accounting in the plant - NUMAC) and Quality control Analyses (Quality of the input bulk chemicals as well as products - QCA). The analytical methods selected are based on the duration of analyses, precision and accuracies required for each type analytical requirement classified earlier. The process and plant control analyses requires lower precision and accuracies as compared to NUMAC analyses, which requires very high precision accuracy. The time taken for analyses should be as lower as possible for process and plant control analyses as compared to NUMAC analyses. The analytical methods required for determining U and Pu in process and plant samples from FRFR will be different as compared to samples from TRFR (Thermal Reactor Fuel Reprocessing) due to higher Pu to U ratio in FRFR as compared TRFR and they should be such that they can be easily

Method of detecting coolant leakages from the pipeways in FBR type reactors

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1986-01-01

Purpose: To detect coolant leakage in the incore pipeways of loop type FBR type reactors in the initial stage at high sensitivity. Constitution: Temperature of the coolants sealed between incore pipeways and the buffle surrounding them is measured by thermocouples and coolant leakage is detected due to fluctuating components. A well-insertion type in which electrode is sealed with argon is used as the thermo-couples. Signals from the thermocouples are once amplified, removed with DC components and then only the fluctuating components are outputted. The fluctuating components are digitalized, passed through an adaptive digital filter and the RMS value as the difference between the output signal and the thermocouple signal is calculated. The calculated value is compared with a threshold value in a comparative calculator. If it exceeds the threshold value, it is judged as abnormal to display an alarm on an alarm display. In this way, the coolant leakage for the pipeways in the FBR type reactor can be detected on real time and at high sensitivity. (Kamimura, M.)

Semi-analytical calculation of fuel parameters for shock ignition fusion

Directory of Open Access Journals (Sweden)

S A Ghasemi

2017-02-01

Full Text Available In this paper, semi-analytical relations of total energy, fuel gain and hot-spot radius in a non-isobaric model have been derived and compared with Schmitt (2010 numerical calculations for shock ignition scenario. in nuclear fusion. Results indicate that the approximations used by Rosen (1983 and Schmitt (2010 for the calculation of burn up fraction have not enough accuracy compared with numerical simulation. Meanwhile, it is shown that the obtained formulas of non-isobaric model cannot determine the model parameters of total energy, fuel gain and hot-spot radius uniquely. Therefore, employing more appropriate approximations, an improved semianalytical relations for non-isobaric model has been presented, which  are in a better agreement with numerical calculations of shock ignition by Schmitt (2010.

Analytical method to evaluate fuel consumption of hybrid electric vehicles at balanced energy content of the electric storage devices

Energy Technology Data Exchange (ETDEWEB)

Katrasnik, Tomaz [University of Ljubljana, Faculty of Mechanical Engineering, Askerceva 6, 1000 Ljubljana (Slovenia)

2010-11-15

Innovative analytically based method to calculate corrected fuel consumption of parallel and series hybrid electric vehicles (HEVs) at balanced energy content of the electric storage devices is proposed and validated in the paper. The proposed analytical method is generally applicable and features highly accurate corrected fuel consumption results. It enables calculation of the corrected fuel consumption out of a single fuel consumption test run in a single analytic post-processing step. An additional fuel consumption test run might be needed to obtain highly accurate results if ratio of the energy content deviation of the electric storage devices to the energy used for vehicle propulsion over the test cycle is high. Proposed method enables consideration of non-linear energy flow changes and non-linear HEV component efficiency changes caused by the energy management strategy or by the component characteristics. The method therefore features highly accurate results out of the minimum number of fuel consumption test runs and thus optimizes workload for development or optimization of HEVs. The input data of the method are characteristic energy flows and efficiencies that are derived from the energy flows on selected energy paths of HEVs. (author)

Comparison of neutron dose measured by Albedo TLD and etched tracks detector at PNC plutonium fuel facilities

International Nuclear Information System (INIS)

Tsujimura, N.; Momose, T.; Shinohara, K.; Ishiguro, H.

1996-01-01

Power Reactor and Nuclear Fuel Development Corporation (PNC) has fabricated Plutonium and Uranium Mixed OXide (MOX) fuel for FBR MONJU at Tokai works. In this site, PNC/Panasonic albedo TLDs/1/ are used for personnel neutron monitoring. And a part of workers wore Etched Tracks Detector (ETD) combined with TLD in order to check the accuracy of the neutron dose estimated by albedo TLD. In this paper, the neutron dose measured by TLD and ETD in the routine monitoring is compared at PNC plutonium fuel facilities. (author)

Fuel Cycle Externalities: Analytical Methods and Issues, Report 2

International Nuclear Information System (INIS)

Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

1994-01-01

that also have not been fully addressed. This document contains two types of papers that seek to fill part of this void. Some of the papers describe analytical methods that can be applied to one of the five steps of the damage function approach. The other papers discuss some of the complex issues that arise in trying to estimate externalities. This report, the second in a series of eight reports, is part of a joint study by the U.S. Department of Energy (DOE) and the Commission of the European Communities (EC)* on the externalities of fuel cycles. Most of the papers in this report were originally written as working papers during the initial phases of this study. The papers provide descriptions of the (non-radiological) atmospheric dispersion modeling that the study uses; reviews much of the relevant literature on ecological and health effects, and on the economic valuation of those impacts; contains several papers on some of the more complex and contentious issues in estimating externalities; and describes a method for depicting the quality of scientific information that a study uses. The analytical methods and issues that this report discusses generally pertain to more than one of the fuel cycles, though not necessarily to all of them. The report is divided into six parts, each one focusing on a different subject area

In pile programme of first valutation of UO2 + PuO2 fuel produced by a new process (GSP)

International Nuclear Information System (INIS)

Caracchin, R.; Lanchi, M.; Marinucci, G.; Nobili, A.; Dupont, G.; Galtier, J.

1982-01-01

The main scope of the ENEA-AGN-CEA programme collaboration is a first valutation of fuel elements produced by GSP method. This valuation will be done by in reactor experiment which enable to compare the performance of GSP and 'standard' FBR fuels. The composition is done by means of theree experimental device: P3, Lugel and Digel. The P3 device gives a direct measurement during irradiation of fuel central temperature, power and integral conductivity. The Lugel device measures fuel stack axial variations and Digel device gives the diameter variations of the pin and PCMI

Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

International Nuclear Information System (INIS)

Kaushik, C.P.

2012-01-01

Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

Spent fuel management in Japan

International Nuclear Information System (INIS)

Mineo, H.; Nomura, Y.; Sakamoto, K.

1998-01-01

In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

Fission product release from nuclear fuel II. Validation of ASTEC/ELSA on analytical and large scale experiments

International Nuclear Information System (INIS)

Brillant, G.; Marchetto, C.; Plumecocq, W.

2013-01-01

Highlights: • A wide range of experiments is presented for the ASTEC/ELSA code validation. • Analytical tests such as AECL, ORNL and VERCORS are considered. • A large-scale experiment, PHEBUS FPT1, is considered. • The good agreement with measurements shows the efficiency of the ASTEC modelling. • Improvements concern the FP release modelling from MOX and high burn-up UO 2 fuels. - Abstract: This article is the second of two articles dedicated to the mechanisms of fission product release from a degraded core. The models of fission product release from nuclear fuel in the ASTEC code have been described in detail in the first part of this work (Brillant et al., this issue). In this contribution, the validation of ELSA, the module of ASTEC that deals with fission product and structural material release from a degraded core, is presented. A large range of experimental tests, with various temperature and conditions for the fuel surrounding atmosphere (oxidising and reducing), is thus simulated with the ASTEC code. The validation database includes several analytical experiments with both bare fuel (e.g. MCE1 experiments) and cladded fuel (e.g. HCE3, VERCORS). Furthermore, the PHEBUS large-scale experiments are used for the validation of ASTEC. The rather satisfactory comparison between ELSA calculations and experimental measurements demonstrates the efficiency of the analytical models to describe fission product release in severe accident conditions

Critical review of analytical techniques for safeguarding the thorium-uranium fuel cycle

International Nuclear Information System (INIS)

Hakkila, E.A.

1978-10-01

Conventional analytical methods applicable to the determination of thorium, uranium, and plutonium in feed, product, and waste streams from reprocessing thorium-based nuclear reactor fuels are reviewed. Separations methods of interest for these analyses are discussed. Recommendations concerning the applicability of various techniques to reprocessing samples are included. 15 tables, 218 references

LBB assessment on ferrite piping structure of large-scale FBR

OpenAIRE

兪 淵植

2002-01-01

These days, this interest on LBB(Leak before Break) design becomes to be rising in the viewpoint of the cost reduction and structural inter-grity for the commercialization of FBR plants, LBB design enables pla-nts to be shut down safely before occuring unstable fracture by dete- cting the leak rates even if a crack initiates and penetrates a wall thickness. It is necessary to assess crack growth and penetration be- havior considering in-service conditions under operation temperature, leak re...

Heat resistant/radiation resistant cable and incore structure test device for FBR type reactor

International Nuclear Information System (INIS)

Tanimoto, Hajime; Shiono, Takeo; Sato, Yoshimi; Ito, Kazumi; Sudo, Shigeaki; Saito, Shin-ichi; Mitsui, Hisayasu.

1995-01-01

A heat resistant/radiation resistant coaxial cable of the present invention comprises an insulation layer, an outer conductor and a protection cover in this order on an inner conductor, in which the insulation layer comprises thermoplastic polyimide. In the same manner, a heat resistant/radiation resistant power cable has an insulation layer comprising thermoplastic polyimide on a conductor, and is provided with a protection cover comprising braid of alamide fibers at the outer circumference of the insulation layer. An incore structure test device for an FBR type reactor comprises the heat resistant/radiation resistant coaxial cable and/or the power cable. The thermoplastic polyimide can be extrusion molded, and has excellent radiation resistant by the extrusion, as well as has high dielectric withstand voltage, good flexibility and electric characteristics at high temperature. The incore structure test device for the FBR type reactor of the present invention comprising such a cable has excellent reliability and durability. (T.M.)

Universal high-temperature heat treatment furnace for FBR mixed uranium and plutonium carbide fuel

International Nuclear Information System (INIS)

Handa, Muneo; Takahashi, Ichiro; Watanabe, Hitoshi

1978-10-01

A universal high-temperature heat treatment furnace for LMFBR advanced fuels was installed in Plutonium Fuel Laboratory, Oarai Research Establishment. Design, construction and performance of the apparatus are described. With the apparatus, heat treatment of the fuel under a controlled gas atmosphere and quenching of the fuel with blowing helium gas are possible. Equipment to measure impurity gas release of the fuel is also provided. Various plutonium enclosure techniques, e.g., a gas line filter with new exchange mechanics, have been developed. In performance test, results of the enclosure techniques are described. (author)

Conceptual design study of LMFBR core with carbide fuel

International Nuclear Information System (INIS)

Tezuka, H.; Hojuyama, T.; Osada, H.; Ishii, T.; Hattori, S.; Nishimura, T.

1987-01-01

Carbide fuel is a hopeful candidate for demonstration FBR(DFBR) fuel from the plant cost reduction point of view. High thermal conductivity and high heavy metal content of carbide fuel lead to high linear heat rate and high breeding ratio. We have analyzed carbide fuel core characteristics and have clarified the concept of carbide fuel core. By survey calculation, we have obtained a correlation map between core parameters and core characteristics. From the map, we have selected a high efficiency core whose features are better than those of an oxide core, and have obtained reactivity coefficients. The core volume and the reactor fuel inventory are approximately 20% smaller, and the burn-up reactivity loss is 50% smaller compared with the oxide fuel core. These results will reduce the capital cost. The core reactivity coefficients are similar to the conventional oxide DFBR's. Therefore the carbide fuel core is regarded as safe as the oxide core. Except neutron fluence, the carbide fuel core has better nuclear features than the oxide core

Investigation of small scale sphere-pac fuel fabrication plant with external gelation process

International Nuclear Information System (INIS)

Maekawa, Kazuhiko; Yoshimura, Tadahiro; Kikuchi, Toshiaki; Hoshino, Yasushi; Munekata, Hideki; Shimizu, Makoto

2005-02-01

In feasibility studies on commercialized FBR cycle system, comprehensive system investigation and properties evaluation for candidate FBR cycle systems have been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation of small scale sphere-pac fuel fabrication plant with external gelation process was conducted. Until last fiscal year, equipment layout in cells and overall layout design of the 200t-HM/y scale fuel fabrication plant were conducted as well as schematical design studies on main equipments in gelation and reagent recovery processes of the plant. System property data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In this fiscal year, the processes from vibropacking to fuel assemblies storage were added to the investigation range, and a conceptual design of whole fuel fabrication plant was studied as well as deepening the design study on main equipments. The conceptual design study was mainly conducted for small 50t-HM/y scale plant and a revising investigation was done for 200t-HM/y scale plant. Taking the planed comparative evaluation with pellet fuel fabrication system into account, design of equipments which should be equivalent with pellet system, especially in post-vibropacking processes, were standardized in each system. Based on these design studies, system properties data concerning economics and environmental burden reduction of the plant was also acquired. In comparison with existing design, the cell height was lowered on condition that plug type pneumatic system was adopted and fuel fabrication building was downsized by applying rationalized layout design of pellet system to post-vibropacking processes. Reduction of reagent usage at gelation process and rationalization of sintering and O/M controlling processes etc., are foremost tasks. (author)

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

International Nuclear Information System (INIS)

Tachibana, Toshimichi

1985-01-01

The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

MONJU experimental data analysis and its feasibility evaluation to build up the standard data base for large FBR nuclear core design

International Nuclear Information System (INIS)

Sugino, K.; Iwai, T.

2006-01-01

MONJU experimental data analysis was performed by using the detailed calculation scheme for fast reactor cores developed in Japan. Subsequently, feasibility of the MONJU integral data was evaluated by the cross-section adjustment technique for the use of FBR nuclear core design. It is concluded that the MONJU integral data is quite valuable for building up the standard data base for large FBR nuclear core design. In addition, it is found that the application of the updated data base has a possibility to considerably improve the prediction accuracy of neutronic parameters for MONJU. (authors)

Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel

International Nuclear Information System (INIS)

Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

2009-01-01

An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

Static analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Dundulis, Gintautas [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania)], E-mail: gintas@mail.lei.lt; Grybenas, Albertas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania); Karalevicius, Renatas [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania); Makarevicius, Vidas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania); Rimkevicius, Sigitas; Uspuras, Eugenijus [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos Street 3, LT-44403 Kaunas (Lithuania)

2009-01-15

The Ignalina Nuclear Power Plant (NPP) has two RBMK-1500 graphite-moderated boiling water multi-channel reactors. The Ignalina NPP Unit 1 was shutdown at the end of 2004, while Unit 2 is foreseen to be shutdown at the end of 2009. At the Ignalina NPP Unit 1 remains approximately 1000 spent fuel assemblies with low burn-up depth. A special set of equipment was developed to reuse these assemblies in the reactor of Unit 2. One of most important items of this set is a container, which is used for the transportation of spent fuel assemblies between the reactors of Unit 1 and Unit 2. A special shock absorber was designed to avoid failure of fuel assemblies in case of hypothetical spent fuel assemblies drop accident during uploading/unloading of spent fuel assemblies to/from container. This shock absorber was examined by using scaled experiments. The objective of this article is the estimation whether the proposed design of shock absorber fulfils the function of the absorber and the optimization of its geometrical parameters using the results of the performed investigations. Static analytical and experimental investigations are presented in the article. The finite element code BRIGADE/Plus was used for the analytical analysis. The calculation model was verified by comparing the experimental investigation and simulation results for further employment of this finite element model in the development of an optimum design of shock absorber. Static simulation was used to perform primary optimization of design and dimension of the shock absorber.

New type fuel exchange system

International Nuclear Information System (INIS)

Meshii, Toshio; Maita, Yasushi; Hirota, Koichi; Kamishima, Yoshio.

1988-01-01

When the reduction of the construction cost of FBRs is considered from the standpoint of the machinery and equipment, to make the size small and to heighten the efficiency are the assigned mission. In order to make a reactor vessel small, it is indispensable to decrease the size of the equipment for fuel exchange installed on the upper part of a core. Mitsubishi Heavy Industries Ltd. carried out the research on the development of a new type fuel exchange system. As for the fuel exchange system for FBRs, it is necessary to change the mode of fuel exchange from that of LWRs, such as handling in the presence of chemically active sodium and inert argon atmosphere covering it and handling under heavy shielding against high radiation. The fuel exchange system for FBRs is composed of a fuel exchanger which inserts, pulls out and transfers fuel and rotary plugs. The mechanism adopted for the new type fuel exchange system that Mitsubishi is developing is explained. The feasibility of the mechanism on the upper part of a core was investigated by water flow test, vibration test and buckling test. The design of the mechanism on the upper part of the core of a demonstration FBR was examined, and the new type fuel exchange system was sufficiently applicable. (Kako, I.)

Method for controlling FBR type reactor

International Nuclear Information System (INIS)

Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

1991-01-01

The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

Back-end fuel cycle efficiencies with respect to improved uranium utilization

International Nuclear Information System (INIS)

Kuczera, B.; Hennies, H.H.

1983-01-01

The world-wide nuclear power plant (NPP) capacity is at present 160 GW(e). If one adds the power stations under construction and ordered, a plant capacity of approximately 480 GW(e) is obtained for 1990, with the share of LWRs making up more than 80%. A modern LWR consumes in the open fuel cycle about 4400 metric tonnes of natural uranium per GW(e), assuming a lifetime of 30 years and a load factor of 70%. Considering the natural uranium reserves known at present and exploitable under economic conditions, it can be conveniently estimated that, with the present NPP capacity extension perspective, the natural uranium resources may be exhausted in a few decades. This trend can be counteracted in a flexible manner by various approaches in fuel cycle technology and strategy: (i) by steady further development of the established LWR technology the uranium consumption can be reduced by about 15%; (ii) closing the nuclear fuel cycle on the basis of LWRs (i.e. thermal uranium and plutonium recycling) implies up to 40% savings in natural uranium consumption; (iii) more recent considerations include the advanced pressurized water reactor (APWR). The APWR combines the proven PWR technology with a newly developed tight lattice core with greatly improved conversion characteristics (conversion ratio = 0.90 to 0.95). In terms of uranium utilization, the APWR has an efficiency three to five times higher than a PWR; (iv) Commercial introduction of FBR systems results in an optimal utilization of uranium which, at the same time, guarantees the supply of nuclear fuel well beyond the present century. For a corresponding transition period an energy supply system can be conceived which relies essentially on extended back-end fuel cycle capacities. These would facilitate a symbiosis of PWR, APWR and FBR, characterized by high flexibility with respect to long-term developments on the energy market. (author)

MOX fuel fabrication: Technical and industrial developments

International Nuclear Information System (INIS)

Lebastard, G.; Bairiot, H.

1990-01-01

The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

FBR structural material test facility in flowing sodium environment

International Nuclear Information System (INIS)

Shanmugasundaram, M.; Kumar, Hemant; Ravi, S.

2016-01-01

In Fast Breeder Reactor (FBR), components such as Control and Safety Rod Drive Mechanism (CSRDM), Diverse Safety Rod Drive Mechanism (DSRDM), Transfer arm and primary sodium pumps etc., are experiencing friction and wear between the moving parts in contact with liquid sodium at high temperature. Hence, it is essential to evaluate the friction and wear behaviour to validate the design of components. In addition, the above core structural reactor components such as core cover plate, control plugs etc., undergoes thermal striping which is random thermal cycling induced by flow stream resulting from the mixing of non isothermal jets near that component. This leads to development of surface cracks and assist in crack growth which in turn may lead to failure of the structural component. Further, high temperature components are often subjected to low cycle fatigue due to temperature gradient induced cyclic thermal stresses caused by start-ups, shutdowns and transients. Also steady state operation at elevated temperature introduces creep and the combination of creep and fatigue leads to creep-fatigue interactions. Therefore, resistance to low cycle fatigue, creep and creep-fatigue are important considerations in the design of FBR components. Liquid sodium is used as coolant and hence the study of the above properties in dynamic sodium are equally important. In view of the above, facility for materials testing in sodium (INSOT) has been constructed and in operation for conducting the experiments such as tribology, thermal stripping, low cycle fatigue, creep and creep-fatigue interaction etc. The salient features of the operation and maintenance of creep and fatigue loops of INSOT facility are discussed in detail. (author)

On recycling of nuclear fuel in Japan

International Nuclear Information System (INIS)

1992-01-01

In Japan, atomic energy has become to accomplish the important role in energy supply. Recently the interest in the protection of global environment heightened, and the anxiety on oil supply has been felt due to the circumstances in Mideast. Therefore, the importance of atomic energy as an energy source for hereafter increased, and the future plan of nuclear fuel recycling in Japan must be promoted on such viewpoint. At present in Japan, the construction of nuclear fuel cycle facilities is in progress in Rokkasho, Aomori Prefecture. The prototype FBR 'Monju' started the general functional test in May, this year. The transport of the plutonium reprocessed in U.K. and France to Japan will be carried out in near future. This report presents the concrete measures of nuclear fuel recycling in Japan from the long term viewpoint up to 2010. The necessity and meaning of nuclear fuel recycling in Japan, the effort related to nuclear nonproliferation, the plan of nuclear fuel recycling for hereafter in Japan, the organization of MOX fuel fabrication in Japan and abroad, the method of utilizing recovered uranium and the reprocessing of spent MOX fuel are described. (K.I.)

Estimating Fuel Cycle Externalities: Analytical Methods and Issues, Report 2

Energy Technology Data Exchange (ETDEWEB)

Barnthouse, L.W.; Cada, G.F.; Cheng, M.-D.; Easterly, C.E.; Kroodsma, R.L.; Lee, R.; Shriner, D.S.; Tolbert, V.R.; Turner, R.S.

1994-07-01

of complex issues that also have not been fully addressed. This document contains two types of papers that seek to fill part of this void. Some of the papers describe analytical methods that can be applied to one of the five steps of the damage function approach. The other papers discuss some of the complex issues that arise in trying to estimate externalities. This report, the second in a series of eight reports, is part of a joint study by the U.S. Department of Energy (DOE) and the Commission of the European Communities (EC)* on the externalities of fuel cycles. Most of the papers in this report were originally written as working papers during the initial phases of this study. The papers provide descriptions of the (non-radiological) atmospheric dispersion modeling that the study uses; reviews much of the relevant literature on ecological and health effects, and on the economic valuation of those impacts; contains several papers on some of the more complex and contentious issues in estimating externalities; and describes a method for depicting the quality of scientific information that a study uses. The analytical methods and issues that this report discusses generally pertain to more than one of the fuel cycles, though not necessarily to all of them. The report is divided into six parts, each one focusing on a different subject area.

Spent fuel treatment in Japan

International Nuclear Information System (INIS)

Takahashi, K.

1999-01-01

In Japan, 52 nuclear power reactors are operating with a total power generation capacity of 45 GWe. The cumulative amount of spent fuel arising, as of March 1998, is about 14,700 W. Spent fuel is reprocessed and recovered nuclear materials are to be recycled in LWRs and FBRs. Pu utilization in LWRs will commence in 1999. In January 1997, short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of the reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, back-end measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away-from-reactor sites, considering the increasing amount of spent fuel arising. Valuable experience was been accumulated at the Tokai Reprocessing Plant (TRP), from the start of hot operation in 1977 up to now. The role of the TRP will be changed from an operation-oriented to a more R and D oriented facility, when PNC is reorganized into the new organization JNC. The Rokkasho reprocessing plant is under construction and is expected to commence operation in 2003. R and D of future recycling technologies is also continued for the establishment of a nuclear fuel cycle based on FBRs and LWRs. (author)

Development of high resolution x-ray CT technique for irradiated fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Asaga, Takeo [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

High X-ray CT technique was developed to observe the irradiation performance of FBR fuel assembly and MOX fuel. In this technique, the high energy X-ray pulse (12MeV) was used synchronizing detection system with the X-ray pulse to reduce the effect of the gamma ray emissions from the irradiated fuel assembly. In this study, this technique was upgraded to obtain high resolution X-ray CT image. In this upgrading, the collimator which had slit width of 0.1 mm and X-ray detector of a highly sensitive silicon semiconductor detector (100 channels) was introduced in the X-ray CT system. As a result of these developments, high resolution X-ray CT images could be obtained on the transverse cross section of irradiated fuel assembly. (author)

Investigation of the growth rate for joint fast breeder reactor and light water reactor operation

International Nuclear Information System (INIS)

Hanan, N.A.; Borg, C.R.; Ott, K.O.

1977-07-01

An investigation of fuel consumption and breeding characteristics of FBR-LWR joint operation is presented. The FBR operates in a closed cycle with joint-reprocessing of core and blanket material. The LWR-portion that runs on FBR plutonium operates in an open cycle. The growth rate of the system is defined based upon the fact that the discharge from the system will make up a fraction of an identical system; the system growth rate is found to have an almost linear dependence on the fraction of the LWR fed by plutonium from the FBR. The LWR growth rate, which is negative, is a constant and represents the fraction of the fuel burnt in the LWR-fraction that runs on FBR plutonium per year

JAEA key facilities for global advanced fuel cycle R and D

Energy Technology Data Exchange (ETDEWEB)

Nomura, Shigeo; Yamamoto, Ryuichi [Nuclear Fuel Cycle Engineering Labos, JAEA, 4-33 Tokai-mura, Ibaraki, 319-1194 (Japan)

2008-07-01

Advanced fuel cycle will be realized with the mid and long term R and D during the long-term transition period from LWR cycle to advanced reactor fuel cycle. Most of JAEA facilities have been utilized to establish the current LWR and FBR (Fast Breeder Reactor) fuel cycle by implementing evolutionary R and D. An assessment of today's state experimental facilities concerning the following research issues: reprocessing, Mox fuel fabrication, irradiation and post-irradiation examination, waste management and nuclear data measurement, is made. The revolutionary R and D requests new issues to be studied: the TRU multi-recycling, minor actinide recycling, the assessment of proliferation resistance and the assessment of cost reduction. To implement the revolutionary R and D for advanced fuel cycle, however, these facilities should be refurbished to install new machines and process equipment to provide more flexible testing parameters.

Vibration-proof FBR type reactor

International Nuclear Information System (INIS)

Kawamura, Yutaka.

1992-01-01

In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

Reprocessing free nuclear fuel production via fusion fission hybrids

Energy Technology Data Exchange (ETDEWEB)

Kotschenreuther, Mike, E-mail: mtk@mail.utexas.edu [Intitute for Fusion Studies, University of Texas at Austin (United States); Valanju, Prashant; Mahajan, Swadesh [Intitute for Fusion Studies, University of Texas at Austin (United States)

2012-05-15

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively 'new' cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th{sup 232}-U{sup 233} conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO{sub 2} matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U{sup 235} fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

Reprocessing free nuclear fuel production via fusion fission hybrids

International Nuclear Information System (INIS)

Kotschenreuther, Mike; Valanju, Prashant; Mahajan, Swadesh

2012-01-01

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th 232 –U 233 conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO 2 matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U 235 fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.

Nuclear power and nuclear weapon proliferation

International Nuclear Information System (INIS)

Apold, A.

1978-01-01

The theme of Dr. Marshall's lecture was that it is, from the viewpoint of prevention of proliferation of nuclear weapons,preferable to use plutonium as a fuel in FBR reactors rather than store it in what, in effect, would be plutonium mines. The true threat of proliferation lies in uranium enrichment. The FBR reactor is misunderstood and the US policy is not against breeders as such. Safeguards against the misuse of plutonium by leaving a residue of radioactivity after reprocessing is quite feasible, despite certain practical problems and extra costs. Weapon proliferation is subject to political objectives and intentions. Definite proposals are, (a) a limited number of reprocessing centres, (b) an accelerated development of FBR reactors, (c) a new FBR fuel cycle, (d) stop storage of spent thermal reactor fuel, (e) reinforced safeguards. (JIW)

Study on Doppler coefficient for metallic fuel fast reactor added hydrogeneous moderator

Energy Technology Data Exchange (ETDEWEB)

Hirakawa, Naohiro; Iwasaki, Tomohiko; Tsujimoto, Kazuhumi [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Osugi, Toshitaka; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Mukaiyama, Takehiko

1998-01-01

A series of mock-up experiments for moderator added metallic fast reactor core was carried out at FCA to obtain the experimental verification for improvement of reactivity coefficients. Softened neutron spectrum increases Doppler effect by a factor of 2, and flatter adjoint neutron spectrum decreases Na void effect by a factor of 0.6 when hydrogen to heavy metal atomic number ratio is increased from 0.02 to 0.13. The experimental results are analyzed with SLALOM and CITATION-FBR, which is the standard design code system for a fast reactor at JAERI, and SRAC95 and CITATION-FBR. The present code system gives generally good agreement with the experimental results, especially by the use of the latter, the dependence of the Doppler effect to the hydrogen to fuel element atomic number density ratio is disappeared. Therefore, it looks possible to use the present code system for the conceptual design of a fast reactor system with hydrogeneous materials. (author)

Developing maintenance technologies for FBR's heat exchanger units by advanced laser processing

International Nuclear Information System (INIS)

Nishimura, Akihiko; Shimada, Yukihiro

2011-01-01

Laser processing technologies were developed for the purpose of maintenance of FBR's heat exchanger units. Ultrashort laser processing fabricated fiber Bragg grating sensor for seismic monitoring. Fiber laser welding with a newly developed robot system repair cracks on inner wall of heat exchanger tubes. Safety operation of the heat exchanger units will be improved by the advanced laser processing technologies. These technologies are expected to be applied to the maintenance for the next generation FBRs. (author)

Training report of the FBR cycle training facility in 2004FY

International Nuclear Information System (INIS)

Watanabe, Toshio; Sasaki, Kazuichi; Sawada, Makoto; Ohtsuka, Jirou

2004-07-01

The FBR cycle training facility consists of sodium handling training facility and maintenance training facility, and is being contributed to train for the operators and maintenance workers of the prototype fast breeder reactor 'Monju'. So far, some training courses have been added to the both training courses of sodium handling technologies maintenance technologies in every year in order to carry out be significant training for preparation of Monju restarting. As encouragement of the sodium handling technology training in 2003FY, the sodium heat transfer basic course was equipped as the 9th sodium handling training course with the aims of learning basic principal technology regarding sodium heat transfer. While, for the maintenance training course, a named 'Monju Systems Learning Training Course', which aims to learn necessary knowledge as the engineers related Monju development, was provided newly in this year as an improvement concerned the maintenance course. In 2003FY, nine sodium handling technology training courses were carried out total 33 times and 235 trainees took part in those training courses. Also, nine training courses concerning the maintenance technology held 15 times and total 113 trainees participated. On the other hand, the 4th special lecture related sodium technology by France sodium school instructor was held on Mar. 15-17 and 34 trainees participated. Consequently, a cumulative trainees since October in 2000 opened the FBR cycle training facility reached to 1,236 so far. (author)

Preliminary Analysis of the Fuel Bundle Stiffness by ANSYS for SFR

Energy Technology Data Exchange (ETDEWEB)

Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-05-15

In SFR (Sodium-cooled Fast Reactor) the temperature of the fuel pin is higher than that of the hexagonal duct, so the thermal expansion rate of the fuel bundle is higher than that of the duct. The neutron fluence and the fuel pin pressure are also increased according to the burnup. So the radial expansion and bowing of a fuel pin bundle would occur, and then fuel bundle would interact with a duct. This phenomenon is called bundle-to-duct interaction (BDI). Under the BDI condition, excess cladding strain and hot spots would occur. Therefore BDI as well as the core mechanics should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE, SHADOW, and MARSE, have been developed to evaluate the BDI behavior. The ANSYS based model is also being developed to analysis the bundle duct interaction for SFR in Korea. In this paper, the fuel pin/bundle model for analyzing the bending deflection and oval deformation was described. The preliminary analysis of the fuel bundle stiffness was performed by the developed model.

FBR Italian position. The PEC reactor and its control rods system

International Nuclear Information System (INIS)

Artioli, C.

1984-01-01

The greatest effort in Italy concerning Fast Breeder Reactors is concentrated on the PEC project. This project represents in fact the sole Italian work on the FBR option. Up to now the state of progress of the PEC project is as follows: - Design work on reactor components and related systems: 80% - Civil engineering work: 45% - Components construction: 18%. It has recently been stated that construction of the plant must be finished by 1987. The progressive design of the PEC control system and the supporting research and development programmes are described in this paper

Development of FBR piping bellows joint

International Nuclear Information System (INIS)

Tsukimori, Kazuyuki; Iwata, Koji

1991-01-01

Reduction of construction cost is one of the most important problems to realize a FBR (Fast Breeder Reactor) Plant. Significant reduction of the construction cost of a reactor building, related equipments and facilities can be expected by shortening the length of its long cooling pipes. Since the bellows has a great capacity for absorbing thermal expansion displacement, application of bellows expansion joints is considered as the most influential measure for reduction of the piping length. To confirm technological possibilities of application and practical use of bellows joints in the main piping systems, extensive R and D's, development of various methods for evaluating the strength of bellows, establishment of inspection and maintenance techniques, studies on safety logic, etc., were carried out by PNC from 1983 to 1988. Through these studies, technological possibilities of bellows joints were confirmed and the results were summarized in the 'Structural Design Guide for Class 1 Piping Bellows Expansion Joints of Fast Breeder Reactor for Elevated Temperature Service' and the 'Inspection and Maintenance Standards of Piping bellows expansion Joints'. (author)

Cleaning and extraction apparatus in a nuclear fuel reprocessing

International Nuclear Information System (INIS)

Nakamura, Yoshiaki.

1983-01-01

Purpose : To eliminate the requirement for the decomposition and cleaning of a centrifugal extractor upon re-processing of FBR type reactor fuels, by preventing solid fission products from depositing on a rotary body of the centrifugal extractor. Constitution : A cleaning and extraction apparatus comprising a combination of a centrifugal cleaner and a centrifugal extractor is used for shortening the contact time between the process liquid and the extraction solvent in FBR type reactor fuel re-processing, and variable parameters are adjusted so that the following equation can be satisfied for avoiding the deposition of solids onto the rotary body of the centrifugal extractor: lsub(e). (rsub(le) 2 + rsub(2r) 2 ) . Nsub(e) . Qsub(c)/ lsub(c) (rsub(lc) 2 + rsub(2c) 2 ) . Nsub(c) . Qsub(e) < 0.8 where Qsub(c) : flow rate to be processed in a centrifugal cleaner, lsub(c) : length of the rotary body, rsub(2c) : radius of a rotary body, rsub(le) : distance from the center to the liquid-extracting hole of the rotary body center to the liquid-extraction hole, Nsub(c) : number of revolution of the rotary body, Qsub(e) : amount of flowrate to be treated in the centrifugal extractor, lsub(e) : length of the rotary body, rsub(2e) : radius for the rotary body, rsub(le) : distance from the center of the rotary body to the liquid discharging aperture and Nsub(e) : number of rotation of the rotary body. (Ikeda, J.)

Nuclear fuels

International Nuclear Information System (INIS)

2008-01-01

-induced damages in structural materials, fuels and targets for the transmutation in FBR; 4 - the fuel of gas-cooled reactors: fuel particulates, behaviour under irradiation, mechanical modeling, the fuel for very high temperature reactors, the fuel for gas-cooled fast reactors; 5 - the fuel for research reactors; 6 - the Jules Horowitz reactor: a tool for the future fuel studies. (J.S.)

Industrial experience of irradiated nuclear fuel reprocessing

International Nuclear Information System (INIS)

Delange, M.

1981-01-01

At the moment and during the next following years, France and La Hague plant particularly, own the greatest amount of industrial experience in the field of reprocessing, since this experience is referred to three types of reactors, either broadly spread all through the world (GCR and LWR) or ready to be greatly developed in the next future (FBR). Then, the description of processes and technologies used now in France, and the examination of the results obtained, on the production or on the security points of view, are a good approach of the actual industrial experience in the field of spent fuel reprocessing. (author)

Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel

International Nuclear Information System (INIS)

Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

2010-01-01

A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

Analytical chemistry instrumentation

International Nuclear Information System (INIS)

Laing, W.R.

1986-01-01

In nine sections, 48 chapters cover 1) analytical chemistry and the environment 2) environmental radiochemistry 3) automated instrumentation 4) advances in analytical mass spectrometry 5) fourier transform spectroscopy 6) analytical chemistry of plutonium 7) nuclear analytical chemistry 8) chemometrics and 9) nuclear fuel technology

Evaluation of alternative fuels for the Greek road transport sector using the analytic hierarchy process

International Nuclear Information System (INIS)

Tsita, Katerina G.; Pilavachi, Petros A.

2012-01-01

This paper evaluates alternative fuels for the Greek road transport sector, using the Analytic Hierarchy Process. Seven different alternatives of fuel mode are considered in this paper: internal combustion engine (ICE) and its combination with petroleum and 1st and 2nd generation biofuels blends, fuel cells, hybrid vehicles, plug-in hybrids and electric vehicles. The evaluation of alternative fuel modes is performed according to cost and policy aspects. In order to evaluate each alternative fuel, one base scenario and ten alternative scenarios with different weight factors selection per criterion are presented. After deciding the alternative fuels’ scoring against each criterion and the criteria weights, their synthesis gives the overall score and ranking for all ten alternative scenarios. It is concluded that ICE blended with 1st and 2nd generation biofuels are the most suitable alternative fuels for the Greek road transport sector. - Highlights: ► Alternative fuels for the Greek road transport sector have been evaluated. ► The method of the AHP was used. ► Seven different alternatives of fuel mode are considered. ► The evaluation is performed according to cost and policy aspects. ► The ICE with 1st and 2nd generation biofuels are the most suitable fuels.

Advanced fuel fabrication

International Nuclear Information System (INIS)

Bernard, H.

1989-01-01

This paper deals with the fabrication of advanced fuels, such as mixed oxides for Pressurized Water Reactors or mixed nitrides for Fast Breeder Reactors. Although an extensive production experience exists for the mixed oxides used in the FBR, important work is still needed to improve the theoretical and technical knowledge of the production route which will be introduced in the future European facility, named Melox, at Marcoule. Recently, the feasibility of nitride fuel fabrication in existing commercial oxide facilities was demonstrated in France. The process, based on carbothermic reduction of oxides with subsequent comminution of the reaction product, cold pressing and sintering provides (U, Pu)N pellets with characteristics suitable for irradiation testing. Two experiments named NIMPHE 1 and 2 fabricated in collaboration with ITU, Karlsruhe, involve 16 nitride and 2 carbide pins, operating at a linear power of 45 and 73 kW/m with a smear density of 75-80% TD and a high burn-up target of 15 at%. These experiments are currently being irradiated in Phenix, at Marcoule. (orig.)

Development of quantitative analytical procedures on two-phase flow in tight-lattice fuel bundles for reduced-moderation light-water reactors

International Nuclear Information System (INIS)

Ohnuki, A.; Kureta, M.; Takae, K.; Tamai, H.; Akimoto, H.; Yoshida, H.

2004-01-01

The research project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR is a light water reactor for which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles whose gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes. (authors)

Liquid-metal fast breeder reactor fuel rod performance and modeling at high burnup

International Nuclear Information System (INIS)

Verbeek, P.; Toebbe, H.; Hoppe, N.; Steinmetz, B.

1978-01-01

The fuel rod modeling codes IAMBUS and COMETHE were used in the analysis and interpretation of postirradiation examination results of mixed-oxide fuel pins. These codes were developed in the framework of the SNR-300 research and development (R and D) program at Interatom and Belgonucleaire, respectively. SNR-300 is a liquid-metal fast breeder reactor demonstration plant designed and presently constructed in consortial cooperation by Germany, Belgium, and the Netherlands. RAPSODIE I, the two-bundle irradiation experiment, was irradiated in the French test FBR RAPSODIE FORTISSIMO and is one of the key irradiation experiments within the SNR-300 R and D program. The comparison of code predictions with postirradiation examination results concentrates on clad diameter expansions, clad total axial elongations, fuel differential and total axial elongations, fuel restructuring, and fission gas release. Fuel rod modeling was considered in the light of benchmarking of the codes, and there was consideration of fuel rod design for operation at low and high burnup

Analytical solutions for the temperature field in a 2D incompressible inviscid flow through a channel with walls of solid fuel

Directory of Open Access Journals (Sweden)

Sorin BERBENTE

2011-12-01

Full Text Available A gas (oxidizer flows between two parallel walls of solid fuel. A combustion is initiated: the solid fuel is vaporized and a diffusive flame occurs. The hot combustion products are submitted both to thermal diffusion and convection. Analytical solutions can be obtained both for the velocity and temperature distributions by considering an equivalent mean temperature where the density and the thermal conductivity are evaluated. The main effects of heat transfer are due to heat convection at the flame. Because the detailed mechanism of the diffusion flame is not introduced the reference chemical reaction is the combustion of premixed fuel with oxidizer in excess. In exchange the analytical solution is used to define an ideal quasi-uniform combustion that could be realized by an n adequate control. The given analytical closed solutions prove themselves flexible enough to adjust the main data of some existing experiments and to suggest new approaches to the problem.

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

International Nuclear Information System (INIS)

Fukaya, Y.; Okubo, T.; Uchikawa, S.

2008-01-01

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

Energy Technology Data Exchange (ETDEWEB)

Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

2008-07-15

The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

Transport and storage of spent nuclear fuel

International Nuclear Information System (INIS)

Lung, M.; Lenail, B.

1987-01-01

From a safety standpoint, spent fuel is clearly not ideal for permanent disposal and reprocessing is the best method of preparing wastes for long-term storage in a repository. Furthermore, the future may demonstrate that some fission products recovered in reprocessing have economic applications. Many countries have in fact reached the point at which the recycling of plutonium and uranium from spent fuel is economical in LWR's. Even in countries where this is not yet evident, (i.e., the United States), the French example shows that the day will come when spent fuel will be retrieved for reprocessing and recycle. It is highly questionable whether spent fuel will ever be considered and treated as waste in the same sense as fission products and processed as such, i.e., packaged in a waste form for permanent disposal. Even when recycled fuel material can no longer be reused in LWR's because of poor reactivity, it will be usable in FBR's. Based on the considerable experience gained by SGN and Cogema, this paper has provided practical discussion and illustrations of spent fuel transport and storage of a very important step in the nuclear fuel management process. The best of spent fuel storage depends on technical, economic and policy considerations. Each design has a role to play and we hope that the above discussion will help clarify certain issues

Interim report of the subcommittee on putting fast breeder reactors in practical use

International Nuclear Information System (INIS)

1983-01-01

Light water reactors are expected to be the main power reactors in Japan for considerably long period hereafter, however, the basic policy of long term energy utilization is to use the plutonium and depleted uranium recovered by the reprocessing of spent fuel for fast breeder reactors, in view of the effective use of energy resources and the energy security. At present, the experimental FBR ''Joyo'' is in operation, and the preparation work was started for the prototype FBR ''Monju'', aiming at the criticality at the end of fiscal 1990. In the discussion, it was confirmed that for the practical use of FBRs, the improvement of their economy is the most important. But the construction cost for the Monju and the type of the demonstration FBR are not yet decided, and there are many uncertain elements. The demand and supply of uranium and the prospect of FBR development, the appropriate way of FBR development in Japan, the trend of FBR development in Japan and foreign countries, the evaluation of the economy of FBRs and the time of putting in practical use, the schedule of FBR development, the cost reduction in FBRs, the fuel cycle and others are reported. (Kako, I.)

Water flow simulation of the flow-induced vibration phenomenon of the thermowell in the prototype-FBR 'Monju'

International Nuclear Information System (INIS)

Anoda, Yoshinari; Kuroda, Takeshi; Kondo, Masaya; Murata, Hideo

1996-06-01

On December 8, 1995 a sodium leak event occurred in the secondary heat transport system (SHTS) of the prototype fast breeder reactor (FBR), Monju, owned and operated by the Power Reactor and Nuclear Fuel Development Corporation (PNC). The direct cause of the leak was a break of a thermowell installed in the loop piping of the SHTS. The break of the thermowell is now believed to have resulted from the flow-induced vibrations due to vortex shedding from the thermowell subjected to a crossflow of sodium. The Japan Atomic Energy Research Institute has conducted a series of water flow model experiments on the flow-induced vibrations of the thermowell to contribute to the post-factor analyses of the event conducted by the Investigation Taskforce on the Sodium Leak Accident in Monju which was established by the Science and Technology Agency (STA) after this event. The experiments were performed for a wide range of experimental conditions including the condition corresponding to the operating condition of the Monju's thermowell and showed the relationship between the vortex shedding pattern and the vibration mode as well as influence of the damping (stability) parameter on the amplitude of vibration. (author)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Development, operational experience and implications for future design of FBRS in Japan

International Nuclear Information System (INIS)

Sawai, S.; Hori, M.

1990-01-01

Joyo, the 100 MW t experimental reactor, has been successfully operated since 1977, and Monju, the 280 MW e prototype FBR, is under construction, with the first criticality planned for 1991. To promote FBR research and development efficiently - including the demonstration FBR (DFBR) programme - a steering committee for R and D was organized in 1986 by the Japan Atomic Power Company, the Power Reactor and Nuclear Fuel Development Corporation, the Japan Atomic Energy Research Institute and the Central Research Institute of Electric Power Industry. A design study of the DFBR is now underway to define its basic specifications by 1990. R and D for Monju, DFBR and future commercial FBRs has been done (1) to improve key technologies developed through the Joyo and Monju programmes; (2) to develop innovative technologies to make FBRs commercial; (3) to promote FBR development in conjunction with the development of the FBR fuel cycle. (author)

BWR noise spectra and application of noise analysis to FBR

International Nuclear Information System (INIS)

Nomura, T.

1975-01-01

Work related to noise analysis, in Tokyo Shibaura Electric Co. Ltd. (Toshiba) and Nippon Atomic Industry Group Co. Ltd. (NAIG) for the past several years is reviewed. After considering the Japan-United States Seminar on Reactor Noise Analysis in 1968, other subjects discussed were boiling water reactor noise analysis and work in relation to FBR. Parts of these are related to each other. For example, boiling detection and temperature fluctuations are problems pertinent to both fields. As the main problems in zero-power-reactor noise are now basically understood, only a brief description of the experiments involving the advanced two detector method is made. Focus is rather placed on the area of power plant noise. (author)

Transport of fresh MOX fuel assemblies for the Monju initial core

International Nuclear Information System (INIS)

Kurakami, J.; Ouchi, Y.; Usami, M.

1997-01-01

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this package design feature such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author)

In-reactor performance of methods to control fuel-cladding chemical interaction

International Nuclear Information System (INIS)

Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G.

1979-01-01

Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (Δ anti GO 2 ) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (Δ anti GO 2 ) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins

Environmental, legal and managerial aspects

International Nuclear Information System (INIS)

1978-09-01

This U.S. contribution to the syllabus for Subgroup 5C treats with environmental and ecological aspects specific to fast breeder reactors, physical protection and safeguarding of the FBR cycle, fuel cycle centers (site selection problems for different degrees of collocation) and administrative and legal problems. Decommissioning of an FBR power plant, syllabus item C.1.7, is treated in separate contribution, more information on advanced safeguards for the fast breeder fyel cycle is contained in the contribution, USA WG 5C-tbd. A key conclusion of this is that with safeguards planning initiated early in the development of the FBR fuel cycle, time is available to develop, evaluate, and implement improved safeguards techniques and incorporate them into the design phase of all FBR cycle facilities

Nuclear power and the possibility of alternative fuel cycles

International Nuclear Information System (INIS)

Engelmann, P.

1979-01-01

Concern about the societal implications, potential risks and the possibility of nuclear weapons proliferation has slowed down the growth of nuclear energy. Assuming a further moderate growth of nuclear power in the Federal Republic of Germany several fuel cycle and reactor strategies can the followed without exhausting the nuclear the resources before the year 2100. The uranium demand of various reactor strategies with LWR's FBR's and HTR's is compared for two demand cases in the FRG. While recycling of spent fuel seems necessary in any case, it is shown that the Th/U cycle can provide a realistic alternative to the U/Pu cycle. The parallel introduction of both cycles appears as the best solution, as it reduces the overall risks and leads to minimum uranium demand. The risk of nuclear proliferation does not vary considerably with the fuel cycle applied; it can, however, be reduced to acceptable levels by safeguards methods and institutional means. (orig.) [de

W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance

International Nuclear Information System (INIS)

Meyers, S.C.; Henderson, J.M.

1980-05-01

The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment

Study on thermalhydraulics of natural circulation decay heat removal in FBR. Experiment with water of typical reactor trip in the demonstration FBR

International Nuclear Information System (INIS)

Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

2010-01-01

Intending to enhance safety and to reduce costs, an FBR plant is being developed in Japan. In relies solely on natural circulation of the primary cooling loop to remove a decay heat of the core after reactor trips. A water test was carried out to advance the development. The test used a 1/10 reduced scale model simulating the core and cooling systems. The experiments simulated representative accidents from steady state to decay heat removal through reactor trip and clarified thermal-hydraulic issues on the thermal circulation performance. Some modifications of the system design were proposed for solving serious problems of natural circulation. An improved design complying with the suggestions will make it possible for natural circulation of the cooling systems to remove the decay heat of the core without causing and unstable or unpredictable change. (author)

Reactor scram device for FBR type reactor

International Nuclear Information System (INIS)

Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

1991-01-01

In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

International Nuclear Information System (INIS)

St John, C.M.

1977-04-01

An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel

Development program of two-arm bilateral servomanipulator system for nuclear fuel cycle facilities in PNC

International Nuclear Information System (INIS)

Kashihara, H.; Igarashi, M.; Maeda, M.; Nomizu, T.; Uematsu, K.

1984-01-01

PNC started the development program of a two-arm bilateral servomanipulator to improve remote maintenance in the planning of pilot plants for HLLW vitrification and FBR fuel reprocessing. In this paper, the development program of two-arm bilateral servomanipulators, signal and power transmissions, high performance viewing system, man-machine interface, and transporter system for manipulators will be presented. Also radiation tests for all components used in a hot cell are being performed to aid remote system design

Integrated Analytic Radionuclide Transport Model for a Spent Nuclear Fuel Repository in Saturated Fractured Rock

International Nuclear Information System (INIS)

Hedin, Allan

2002-01-01

Simple analytic expressions are presented for radionuclide transport from a KBS 3-type repository, where spent nuclear fuel is placed in copper canisters surrounded by bentonite clay and deposited at a depth of 500 m in fractured granitic rock.Dissolution of readily accessible and fuel matrix embedded nuclides, chain decay, and nuclide precipitation is treated within the canister. Transport in the canister void and buffer is modeled with a dual stirred tank analogy, where transport resistances represent an assumed small initial damage in the canister and transport features of the buffer-geosphere interface. Initial, transient diffusion in the buffer is treated with a simple correction term. Chain decay is not included in the buffer.Geosphere transport expressions handle advection, longitudinal dispersion, matrix diffusion, sorption, and radioactive decay, but not chain decay. The treatment is based on earlier results for an instantaneous inlet and for a constant inlet to the geosphere in the nondispersive case. A correction is added so that longitudinal dispersion is taken approximately into account. The correction utilizes analytical expressions for the temporal moments of the geosphere release curve in the dispersive case.The near-field/geosphere integration is treated in a simplified manner avoiding numerical convolutions. The instantaneous inlet expression for the geosphere release is used when the near-field release decreases rapidly in comparison to a typical response time in the geosphere; the constant inlet expression is used in the opposite case.Twenty-seven calculation cases from a safety assessment of a KBS 3 repository using borehole data from three different field investigation sites were repeated with the analytic expressions. The agreement in both near-field and geosphere releases is in general well within an order of magnitude for the variety of long- and short-lived, sorbing, nonsorbing, solubility limited, immediately accessible, and fuel matrix

CANDU advanced fuel cycles

International Nuclear Information System (INIS)

Slater, J.B.

1986-03-01

This report is based on informal lectures and presentations made on CANDU Advanced Fuel Cycles over the past year or so, and discusses the future role of CANDU in the changing environment for the Canadian and international nuclear power industry. The changing perspectives of the past decade lead to the conclusion that a significant future market for a CANDU advanced thermal reactor will exist for many decades. Such a reactor could operate in a stand-alone strategy or integrate with a mixed CANDU-LWR or CANDU-FBR strategy. The consistent design focus of CANDU on enhanced efficiency of resource utilization combined with a simple technology to achieve economic targets, will provide sufficient flexibility to maintain CANDU as a viable power producer for both the medium- and long-term future

FBR type reactors

International Nuclear Information System (INIS)

Kawashima, Katsuyuki; Azekura, Kazuo; Inoue, Kotaro.

1981-01-01

Purpose: To decrease power fluctuations due to burning of blanket fuel element clusters by partially replacing the fertile materials in the blanket fuel element clusters with fissile materials. Constitution: Fertile materials in the radial blanket fuel element clusters disposed to the outside or inside of the reactor core are partially replaced with fissile materials. Since the power density of the fissile materials is at the maximum in the initial burning stage and decreases as the burning proceeds, the power density of the materials which is smaller in the initial burning stage and becomes greater with the burning by the neutron-accumulated plutonium is offset. Accordingly, the power fluctuations in the blanket fuel element clusters due to the burning made smaller thereby enable to form a reactor core with less power fluctuations due to burning under the constant coolant flow rate depending on the power in the final burning stage where the blanket power is maximum. (Moriyama, K.)

After-heat removing system in FBR type reactor

International Nuclear Information System (INIS)

Ohashi, Yukio.

1990-01-01

The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

Reduction of HLW repository environmental impact by Na-Cooled FBR

International Nuclear Information System (INIS)

Ahn, Joonhong; Ikegami, Tetsuo

2005-01-01

The present study proposes the total toxicity index of radionuclides that have accumulated in the region exterior to the repository as the environmental impact measure. The measure is quantitatively evaluated by a radionuclide transport models that incorporate the effects of canister-array configuration and the initial mass loading in the waste canister. With the measure, it is demonstrated that the environmental impact of the repository can be effectively reduced by reduction of the initial mass loading resulting from FBR deployment and change in the canister-array configuration in the repository. The rate of increase in the environmental impact with the increase in the repository size can be reduced by reducing the initial mass loading of Np and its precursors. (author)

Development of thermal hydraulic analysis code for IHX of FBR

International Nuclear Information System (INIS)

Kumagai, Hiromichi; Naohara, Nobuyuki

1991-01-01

In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)

Studies on use of reflector material and its position within FBR core for reducing U{sup 232} content of U produced in ThO{sub 2} radial blankets

Energy Technology Data Exchange (ETDEWEB)

Sen, Sujoy, E-mail: sujoy@igcar.gov.in [Core Design Group, IGCAR, Kalpakkam (India); Prasad, Rajeev Ranjan; Bagchi, Subhrojit [Core Design Group, IGCAR, Kalpakkam (India); Mohanakrishnan, P. [MCNS, Manipal University, Manipal (India); Arul, A. John; Puthiyavinayagam, P. [Core Design Group, IGCAR, Kalpakkam (India)

2015-11-15

Highlights: • Nuclear data processing for multigroup neutron transport calculation. • Discrete ordinate and Monte Carlo neutron transport. • Breeding of Thorium in Fast Reactor. • Minimization of U{sup 232} in U{sup 233}. • Fuel burn up using Neutron Diffusion. - Abstract: Presence of U{sup 232} in U{sup 233} bred in thorium blanket of fast reactor is a major concern in fuel reprocessing. The former's daughter products being hard gamma emitter and the isotope itself having substantial half life, its presence beyond 10 ppm makes fuel recycle complicated and expensive. In this study possibility of decreasing U{sup 232} production in a typical FBR blanket by means of spectrum modification is examined. SS, depleted B{sub 4}C, SiC, Mo and W regions were introduced between core and radial blanket and evolution of isotopes were studied to arrive at an optimal configuration that satisfies requirements of breeding U{sup 233} and lowering U{sup 232}concentration. SS, B{sub 4}C, SiC, Mo and W are known to be high temperature material with appropriate stability in harsh fast reactor environment. Study has shown that introducing two SS reflector rows can achieve the required low value of U{sup 232}concentration without greatly compromising the U{sup 233}production.

Development of dose rate estimation system for FBR maintenance

Energy Technology Data Exchange (ETDEWEB)

Iizawa, Katsuyuki [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Takeuchi, Jun; Yoshikawa, Satoru [Hitachi Engineering Company, Ltd., Hitachi, Ibaraki (Japan); Urushihara, Hiroshi [Ibaraki Hitachi Information Service Co., Ltd., Omika, Ibaraki (Japan)

2001-09-01

During maintenance activities on the primary sodium cooling system by an FBR Personnel radiation exposure arises mainly from the presence of radioactive corrosion products (CP). A CP behavior analysis code, PSYCHE, and a radiation shielding calculation code, QAD-CG, have been developed and applied to investigate the possible reduction of radiation exposure of workers. In order to make these evaluation methods more accessible to plant engineers, the user interface of the codes has been improved and an integrated system, including visualization of the calculated gamma-ray radiation dose-rate map, has been developed. The system has been verified by evaluating the distribution of the radiation dose-rate within the Monju primary heat transport system cells from the estimated saturated CP deposition and distribution which would be present following about 20 cycles of full power operation. (author)

Development of dose rate estimation system for FBR maintenance

International Nuclear Information System (INIS)

Iizawa, Katsuyuki; Takeuchi, Jun; Yoshikawa, Satoru; Urushihara, Hiroshi

2001-01-01

During maintenance activities on the primary sodium cooling system by an FBR Personnel radiation exposure arises mainly from the presence of radioactive corrosion products (CP). A CP behavior analysis code, PSYCHE, and a radiation shielding calculation code, QAD-CG, have been developed and applied to investigate the possible reduction of radiation exposure of workers. In order to make these evaluation methods more accessible to plant engineers, the user interface of the codes has been improved and an integrated system, including visualization of the calculated gamma-ray radiation dose-rate map, has been developed. The system has been verified by evaluating the distribution of the radiation dose-rate within the Monju primary heat transport system cells from the estimated saturated CP deposition and distribution which would be present following about 20 cycles of full power operation. (author)

Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

International Nuclear Information System (INIS)

Clement Ravi Chandar, S.; Sivayya, D.N.; Puthiyavinayagam, P.; Chellapandi, P.

2013-01-01

For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B 4 C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability

Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

Mid-Term Direction of JAEA Nuclear Fuel Cycle Engineering Laboratories

International Nuclear Information System (INIS)

Ojima, H.; Sugiyama, T.; Tanaka, K.; Takeda, S.; Nomura, S.

2009-01-01

1. Introduction Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) has sufficient experience and ability through its 50 year operation to establish the next generation closed cycle. It strives to become a world-class Center Of Excellence. 2. Current activity in NCL: 1) - Recycling of MOX fuel: The Tokai Reprocessing Plant has reprocessed 29 tons of MOX fuel from the ATR Fugenh as a part of 1140 tons of cumulative spent fuel reprocessed. JAEA has supported the pre-operation of the Rokkasho Reprocessing Plant. An innovative MOX pellet fabrication process has been developed in the Plutonium Fuel Development Center, and a part of products obtained by the development are used as a fuel for core confirmation test for re-startup of the FBR Monjuh. Characterization of MOX containing Am and Np has been studied and a new data such as melting point and thermal conductivity were reported. In the Chemical Processing Facility, a hot lab., an advanced aqueous reprocessing technology has been tested for TRU recovery, economical improvement, etc., using irradiated MOX fuel from the FR Joyoh. 2) - Supporting Activity: JAEA has improved the effectiveness and efficiency of existing safeguards activities. The Integrated Safeguards approach for all facilities in NCL has been implemented since August, 2008, as a pioneer and a good example in the world. To reduce anxiety among local residents, NCL has explained its operation plans and exchanged information and opinions with them concerning potential risks to health and environment. Recently, stake-holder participation in the management of NCL was started from the view point of Corporate Social Responsibility. In April, 2008, the agreement was signed with Idaho National Laboratory for cooperation of personnel training in fuel cycle area. 3. Mid-Term Direction: In Japan, feasibility and direction of the transition period from the LWR era to the FBR era should be discussed for the next several years. Study

Mid-Term Direction of JAEA Nuclear Fuel Cycle Engineering Laboratories

Energy Technology Data Exchange (ETDEWEB)

Ojima, H.; Sugiyama, T.; Tanaka, K.; Takeda, S.; Nomura, S. [Tokai-mura, Ibaraki-ken 319-1194 (Japan)

2009-06-15

1. Introduction Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) has sufficient experience and ability through its 50 year operation to establish the next generation closed cycle. It strives to become a world-class Center Of Excellence. 2. Current activity in NCL: 1) - Recycling of MOX fuel: The Tokai Reprocessing Plant has reprocessed 29 tons of MOX fuel from the ATR Fugenh as a part of 1140 tons of cumulative spent fuel reprocessed. JAEA has supported the pre-operation of the Rokkasho Reprocessing Plant. An innovative MOX pellet fabrication process has been developed in the Plutonium Fuel Development Center, and a part of products obtained by the development are used as a fuel for core confirmation test for re-startup of the FBR Monjuh. Characterization of MOX containing Am and Np has been studied and a new data such as melting point and thermal conductivity were reported. In the Chemical Processing Facility, a hot lab., an advanced aqueous reprocessing technology has been tested for TRU recovery, economical improvement, etc., using irradiated MOX fuel from the FR Joyoh. 2) - Supporting Activity: JAEA has improved the effectiveness and efficiency of existing safeguards activities. The Integrated Safeguards approach for all facilities in NCL has been implemented since August, 2008, as a pioneer and a good example in the world. To reduce anxiety among local residents, NCL has explained its operation plans and exchanged information and opinions with them concerning potential risks to health and environment. Recently, stake-holder participation in the management of NCL was started from the view point of Corporate Social Responsibility. In April, 2008, the agreement was signed with Idaho National Laboratory for cooperation of personnel training in fuel cycle area. 3. Mid-Term Direction: In Japan, feasibility and direction of the transition period from the LWR era to the FBR era should be discussed for the next several years. Study

Steam generator of FBR type reactor

International Nuclear Information System (INIS)

Hashiguchi, Ko.

1992-01-01

Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

JAEA FBR Plant Engineering Center annual report 2011

International Nuclear Information System (INIS)

2012-11-01

The FBR Plant Engineering Center was established on April 1, 2009 located in a research building, of which care is taken by the International Nuclear Information Training Center, Tsuruga Head Office, at Shiraki in Tsuruga. The mission of the center is to perform R and D (research and development) works both for analysis of operational experiences at the prototype fast breeder reactor “Monju” and for technology development concerning design and operation of “Monju”. Moreover it is also required to apply the results to next generation fast breeder reactors, which is an important role of Advanced Nuclear System Research and Development Directorate. And in these R and D activities, it is expected to conduct the works in cooperation with domestic or foreign research organizations or universities by a joint-study or a collaborative-work manner. The R and D activities have been carried out specifically on the “demonstration of the reliability as a power generation plant” and “establishment of sodium handling technology”, which are originally intended missions of “Monju”. And the other R and Ds have been promoted both for the plant engineering, such as plant maintenance, to effectively use an existing reactor in order to apply the R and D results to a future demonstration reactor, and for the irradiation test study, such as advanced fuel irradiation, to use “Monju” as an irradiation test bed. In order to perform these R and D activities, five R and D groups have been set up in the center. They are operation-and-maintenance engineering, sodium engineering, reactor-core-and-fuel engineering, plant engineering, and safety engineering groups. However, the Japanese atomic energy policy is being reviewed after the accident of the Fukushima Daiichi nuclear power station caused by a tsunami generated by the Tohoku-district-off-the-Pacific-Ocean Earthquake on March 11, 2011, and all the R and D activities using “Monju” have been suspended since late 2011

Recent R/D towards aqueous reprocessing of FBR fuels

Energy Technology Data Exchange (ETDEWEB)

Mallika, C.; Pandey, N.K.; Kumar, S.; Kamachi Mudali, U. [Materials, Process and Equipment Development Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

2016-07-01

The mixed Pu-rich carbide spent fuel with a burn up of 155 GWd/t from the Fast Breeder Test Reactor is being reprocessed in a hot-cell facility by PUREX process. Based on the input from the operation of this facility, research and development activities were carried out to improve the recovery, decontamination factors, economy and to reduce the waste volumes. Reduction of uranyl ions in a continuous flow electrochemical reactor and electrolytic as well as chemical reduction of 4 M HNO{sub 3} from liquid waste could be performed in continuous mode. Using the optimized parameters, suitable electrolytic cells/experimental setups were designed for the plant capacity of 6 L/h. Studies on the extraction kinetics of Ru with 30% TBP (tributyl phosphate) in NPH revealed that better decontamination factor with respect to Ru can be achieved using fast contactors like centrifugal extractors (CEs). Towards developing a spent solvent recovery system to reduce organic waste volumes, a pilot plant was set up, which could recover diluent as top product of distillation column and 40% TBP as bottom product from inactive degraded solvent. A solvent recovery system using short path distillation was also developed for installation in hot cells. (authors)

A new concept of hydrogen production system for sodium cooled FBR

International Nuclear Information System (INIS)

Nakagiri, Toshio; Aoto, Kazumi; Hoshiya, Taiji

2004-01-01

A new thermo-chemical and electrolytic hybrid hydrogen production process (thermo-chemical and electrolytic Hybrid Hydrogen process in Lower Temperature range: HHLT) is newly proposed by the Japan Nuclear Cycle Development Institute (JNC) to realize the hydrogen production from water by using the heat generation of sodium cooled Fast Breeding Reactor (FBR). The HHLT process is based on the sulfuric acid (H 2 SO 4 ) synthesis and decomposition processes developed earlier (Westinghouse process), and sulfur trioxide (SO 3 ) decomposition process of HHLT is facilitated by electrolysis with ionic oxygen conductive solid electrolyte to reduce operating temperature 200degC-300degC lower than Westinghouse process. Decomposition processes of SO 3 were confirmed with the cell voltage lower than 0.5 V at 500degC-600degC using 8mol yttria stabilized zirconia (8molYSZ) solid electrolyte and platinum electrode. Therefore, total voltage required for HHLT is expected to be lower than 1.0 V, because the voltage required for sulfuric acid synthesis is about 0.5V. Thermal efficiency of HHLT based on chemical reactions was roughly estimated to be within the range of 35% to 55% under the influence of H 2 SO 4 concentration and heat recovery. These results show the possibility of development of a new hydrogen production process which needs low splitting voltage and has high efficiency at around 500degC, utilizing the heat generation of sodium cooled FBR. SO 3 splitting with the voltage lower than 0.5V was confirmed at about 500degC experimentally, and ideal thermal efficiency of the cycle based on chemical reactions was evaluated. Furthermore, test apparatus to substantiate whole process of HHLT was manufactured. (author)

Long-term fuel cycle scenarios for advanced utilization of plutonium from LWRs

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji

2005-01-01

The Innovative Water Reactor for Flexible fuel cycle (FLWR) realizes multiple recycling and breeding of Pu, which enables effective utilization of the uranium resource, and is based on well-developed LWR technologies. This reactor offers flexibility for the future nuclear fuel cycle situation. Three scenarios were defined for future deployment of nuclear power generation and fuel cycle systems in Japan and analyzed from the view point of Pu recycle, natural uranium consumption and stock of spent fuels. The LWR with long-term Pu recycle with or without MOX fuel reprocessing needs uranium of about 9 thousands tons per year and accumulated uranium consumption of 1.5 million tons in 2150. If the FLWR with net conversion ratio of 0.89 and 1.04 would be introduced in 2025 and 2050 or 2030, it would suppress ultimate required natural uranium and control the uranium consumption about less than 1.2 million tons in 2150, while the FLWR in 2025 and FBR with breeding ratio of 1.16 in 2050 will at 0.9 million tons after in 2100. (T. Tanaka)

Development of the adjusted nuclear cross-section library based on JENDL-3.2 for large FBR

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki

1999-04-01

JNC (and PNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In 1991, the adjusted library based on JENDL-2, JFS-3-J2 (ADJ91R), was developed, and it has been used on the design research for FBR. As an evaluated nuclear library, however, JENDL-3.2 is recently used. Therefore, the authors developed an adjusted library based on JENDL-3.2 which is called JFS-3-J3.2(ADJ98). It is known that the adjusted library based on JENDL-2 overestimated the sodium void reactivity worth by 10-20%. It is expected that the adjusted library based on JENDL-3.2 solve the problem. The adjusted library JFS-3-J3.2(ADJ98) was produced with the same method as the adjusted library JFS-3-J2(ADJ91R) and used more integral parameters of JUPITER experiments than the adjusted library JFS-3-J2(ADJ91R). This report also describes the design accuracy estimation on a 600 MWe class FBR with the adjusted library JFS-3-J3.2(ADJ98). Its main nuclear design parameters (multiplication factor, burn-up reactivity loss, breeding ratio, etc.) except the sodium void reactivity worth which are calculated with the adjusted library JFS-3-J3.2(ADJ98) are almost the same as those predicted with JFS-3-J2(ADJ91R). As for the sodium void reactivity, the adjusted library JFS-3-J3.2(ADJ98) estimates about 4% smaller than the JFS-3-J2(ADJ91R) because of the change of the basic nuclear library from JENDL-2 to JENDL-3.2. (author)

Investigation of safety measures to severe accident of Fast Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

So as to plan the accident management to severe accident of Fast Breeder Reactor (FBR), it is primary important to understand the progression of severe accident (SA) precisely. In this study, it has been aimed to reveal two items that work as keys in the evaluation of SA in sodium cooled FBR. One is the cool-ability of degraded core on the core support plate by sodium natural circulation in the post accident heat removal (PAHR) phase. An obstacle that hinders the smooth heat transfer from fuel debris to coolant is the formation of sodium-uranate by chemical reaction between sodium and fuel. Following the measurement of physical values of sodium-uranate in FY 2011, experiments has been performed to reveal the conditions for sodium-uranate formation on fuel debris in sodium pool simulating the actual situation of the degraded core. The cool-ability of the debris bed was analyzed using the Lipinski 1-D model. Another research performed in this study is the measurement of fission product (cesium and antimony) evaporation rates from FBR fuel as a function of temperature, because presently the fission product evaporation rates data for LWR is also temporarily used for FBR SA analysis. The measurement was performed using the irradiated fuels in the Test Reactor JOYO. (author)

An analytical model and parametric study of electrical contact resistance in proton exchange membrane fuel cells

Energy Technology Data Exchange (ETDEWEB)

Wu, Zhiliang; Wang, Shuxin; Zhang, Lianhong [School of Mechanical Engineering, Tianjin University, Tianjin 300072 (China); Hu, S. Jack [Department of Mechanical Engineering, The University of Michigan, Ann Arbor, MI 48109-2125 (United States)

2009-04-15

This paper presents an analytical model of the electrical contact resistance between the carbon paper gas diffusion layers (GDLs) and the graphite bipolar plates (BPPs) in a proton exchange membrane (PEM) fuel cell. The model is developed based on the classical statistical contact theory for a PEM fuel cell, using the same probability distributions of the GDL structure and BPP surface profile as previously described in Wu et al. [Z. Wu, Y. Zhou, G. Lin, S. Wang, S.J. Hu, J. Power Sources 182 (2008) 265-269] and Zhou et al. [Y. Zhou, G. Lin, A.J. Shih, S.J. Hu, J. Power Sources 163 (2007) 777-783]. Results show that estimates of the contact resistance compare favorably with experimental data by Zhou et al. [Y. Zhou, G. Lin, A.J. Shih, S.J. Hu, J. Power Sources 163 (2007) 777-783]. Factors affecting the contact behavior are systematically studied using the analytical model, including the material properties of the two contact bodies and factors arising from the manufacturing processes. The transverse Young's modulus of chopped carbon fibers in the GDL and the surface profile of the BPP are found to be significant to the contact resistance. The factor study also sheds light on the manufacturing requirements of carbon fiber GDLs for a better contact performance in PEM fuel cells. (author)

Nuclear symbiosis. Why, when and how

International Nuclear Information System (INIS)

Stewart, H.B.

1978-01-01

A distinction can be made between an ''asymptotic'' nuclear symbiosis and a ''transitional'' nuclear symbiosis. With the currently projected nuclear energy growth, there appears to be no urgency for introducing nuclear systems required for the asymptotic era before, say, another 50 to 75 years. Consequently, more attention should be directed toward desirable systems for the transitional era, i.e., the next 30 to 50 years. In the classical asymptotic system, an optimal system would consist of a relatively small number of fast breeder reactor plants operating largely as ''fuel factories'' and a large number of thermal spectrum near breeder reactor plants serving as ''energy factories''. The breeding ratio of 238 U/Pu in the FBR plants would be chosen at unity to assure that FBR primary fuel could sustain itself. Excess breeding capacity would be utilized to convert 232 Th and 233 U in the blankets. This excess fuel would then be used to supply make-up requirements for thermal near breeder (NBR) plants. With an FBR breeder ratio of 1.4 and an NBR conversion ratio of 0.9, approximately four NBR plants could be supported for each FBR plant. Perhaps more importantly, the sale of 233 U to the NBR could be used to offset a probable higher capital cost of the FBR plants. In this way, a capital cost penalty of some 30 to 40% relative to LWR plants might be allowable for FBR plants. (author)

Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels

International Nuclear Information System (INIS)

Talamo, Alberto

2007-01-01

We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in 235 U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides 240 Pu, 238 U and 232 Th and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for 240 Pu, 238 U and 232 Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 μm and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core

Development and evaluation of analytical techniques for total chlorine in used oils and oil fuels

International Nuclear Information System (INIS)

Gaskill, A. Jr.; Estes, E.D.; Hardison, D.L.; Friedman, P.H.

1990-01-01

A current EPA regulation prohibits the sale for burning in nonindustrial boilers of used oils and oil fuels. This paper discusses how analytical techniques for determining total chlorine were evaluated to provide regulatory agencies and the regulated community with appropriate chlorine test methods. The techniques evaluated included oxygen bomb combustion followed by chemical titration or ion chromatography, instrumental microcoulometry, field test kits, and instrumental furnace/specific ion electrode determinator, a device based on the Beilstein reaction, and x-ray fluorescence spectrometry. These techniques were subjected to interlaboratory testing to estimate their precision, accuracy, and sensitivity. Virgin and used crankcase oils, hydraulic and metalworking oils, oil fuels and oil fuel blends with used oils were tested. The bomb techniques, one of the test kits, microcoulometry and all but one x-ray analyzer were found to be suitable for this application. The chlorine furnace and the Beilstein device were found to be inapplicable at the levels of interest

China's spent fuel treatment: The present status and prospects

International Nuclear Information System (INIS)

Jiang Yunqing

1999-01-01

In the mid 1980s, China launched the development of nuclear power dominated by PWRs and opted for the closed fuel cycle strategy. On the basis of irradiated fuel reprocessing for defence purpose, an R and D programme for civil reprocessing has been implemented. Currently, China's spent fuel arising is limited but its amount will sharply increase with nuclear power expansion early next century. Spent fuel stored at reactor site for at least 5 years will be transported either by a combination of sea and rail or by road directly to the Lanzhou Nuclear Fuel Complex. A wet centralized storage facility with a 550 tHM capacity has been built for interim storage of spent fuel. Also, a multi-purpose reprocessing pilot plant with a maximum throughput of 400 kg HM/d is now under construction and will be put into commissioning by the turn of the century. A large-scale commercial reprocessing plant, perhaps with a capacity of 800 tHM/a, will be set up around 2020. Recovered uranium and plutonium from reprocessing will go to a demonstration plant and be manufactured into MOX fuel for FBR and PWR. The defence radwaste from reprocessing is at present being conditioned into the proper forms and will be disposed in appropriate repositories. All expertise and experience gained from these practices will be utilized in the future civil radwaste management. (author)

Evaluation of creep-fatigue crack growth for large-scale FBR reactor vessel and NDE assessment

Energy Technology Data Exchange (ETDEWEB)

Joo, Young Sang; Kim, Jong Bum; Kim, Seok Hun; Yoo, Bong

2001-03-01

Creep fatigue crack growth contributes to the failure of FRB reactor vessels in high temperature condition. In the design stage of reactor vessel, crack growth evaluation is very important to ensure the structural safety and setup the in-service inspection strategy. In this study, creep-fatigue crack growth evaluation has been performed for the semi-elliptical surface cracks subjected to thermal loading. The thermal stress analysis of a large-scale FBR reactor vessel has been carried out for the load conditions. The distributions of axial, radial, hoop, and Von Mises stresses were obtained for the loading conditions. At the maximum point of the axial and hoop stress, the longitudinal and circumferential surface cracks (i.e. PTS crack, NDE short crack and shallow long crack) were postulated. Using the maximum and minimum values of stresses, the creep-fatigue crack growth of the proposed cracks was simulated. The crack growth rate of circumferential cracks becomes greater than that of longitudinal cracks. The total crack growth of the largest PTS crack is very small after 427 cycles. The structural integrity of a large-scale reactor can be maintained for the plant life. The crack depth growth of the shallow long crack is faster than that of the NDE short crack. In the ISI of the large-scale FBR reactor vessel, the ultrasonic inspection is beneficial to detect the shallow circumferential cracks.

Accelerator breeder: a viable option for the production of nuclear fuels

International Nuclear Information System (INIS)

Grand, P.

1983-01-01

Despite the growing pains of the US nuclear power industry, our dependence on nuclear energy for the production of electricity and possibly process heat is likely to increase dramatically over the next few deacades. This statement dismisses fusion as being entirely too speculative to be practical within that time frame. Sometime, between the years 2000 and 2050, fissile material will be in short supply whether it is to fuel existing LWR's or to provide initial fuel inventory for FBR's. The accelerator breeder could produce the fuel shortfall predicted to occur during the first half of the 21st century. The accelerator breeder offers the only practical means today of producing, or breeding, large quantities of fissile fuel from fertile materials, albeit at high cost. Studies performed over the last few years at Chalk River Laboratory and at Brookhaven National Laboratory have demonstrated that the accelerator breeder is practical, technically feasible with state-of-the-art technology, and is economically competitive with any other proposed synthetic means of fissile fuel production. This paper gives the parameters of a nearly optimized accelerator-breeder system, then discusses the development needs, and the economics and institutional problems that this breeding concept faces

Reactor core for FBR type reactor

International Nuclear Information System (INIS)

Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

1996-01-01

In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

Analytical Chemistry Department annual report, 1975

International Nuclear Information System (INIS)

Mosen, A.W.

1976-01-01

The analytical methods developed or adopted for use in support of radiochemistry and gamma ray spectroscopy, HTGR fuel reprocessing, HTGR fuel development, TRIGA fuel fabrication, and miscellaneous projects are reported

Analytical methods for fissionable materials in the nuclear fuel cycle. Covering June 1974--June 1975

International Nuclear Information System (INIS)

Waterbury, G.R.

1975-10-01

Research progress is reported on method development for the dissolution of difficult-to-dissolve materials, the automated analysis of plutonium and uranium, the preparation of plutonium materials for the Safeguard Analytical Laboratory Evaluation (SALE) Program, and the analysis of HTGR fuel and SALE uranium materials. The previously developed Teflon-container, metal-shell apparatus was applied to the dissolution of various nuclear materials. Gas--solid reactions, mainly using chlorine at elevated temperatures, are promising for separating uranium from refractory compounds. An automated spectrophotometer designed for determining plutonium and uranium was tested successfully. Procedures were developed for this instrument to analyze uranium--plutonium mixtures and the effects of diverse ions upon the analysis of plutonium and uranium were further established. A versatile apparatus was assembled to develop electrotitrimetric methods that will serve as the basis for precise automated determinations of plutonium. Plutonium materials prepared for the Safeguard Analytical Laboratory Evaluation (SALE) Program were plutonium oxide, uranium--plutonium mixed oxide, and plutonium metal. Improvements were made in the methods used for determining uranium in HTGR fuel materials and SALE uranium materials. Plutonium metal samples were prepared, characterized, and distributed, and half-life measurements were in progress as part of an inter-ERDA-laboratory program to measure accurately the half-lives of long-lived plutonium isotopes

Leak detector for a steam generator in FBR type reactors

International Nuclear Information System (INIS)

Miyaji, Nobuyoshi.

1979-01-01

Purpose: To facilitate maintenance for liquid leak detectors such as exchange of nickel membrane sensors during operation in a sodium-cooled fbr type reactor. Constitution: A pipeway capable of supplying a cover gas such as argon into the cylinder of a hydrogen detector containing a nickel membrane sensor is provided in a liquid leak detector constituting a part of a by-pass loop. The pipeway is also adapted to be evacuated. A pipeway and a small sodium tank for drain use are provided on the side of the by-pass loop near valves. Then, after closing the inlet and outlet valves to disconnect the by-pass loop from the sodium main pipeway, the cover gas is supplied to drive liquid sodium to the drain tank. After the drain of the liquid sodium, the sensor can be replaced. (Ikeda, J.)

Analytical calculation of electrolyte water content of a Proton Exchange Membrane Fuel Cell for on-board modelling applications

Science.gov (United States)

Ferrara, Alessandro; Polverino, Pierpaolo; Pianese, Cesare

2018-06-01

This paper proposes an analytical model of the water content of the electrolyte of a Proton Exchange Membrane Fuel Cell. The model is designed by accounting for several simplifying assumptions, which make the model suitable for on-board/online water management applications, while ensuring a good accuracy of the considered phenomena, with respect to advanced numerical solutions. The achieved analytical solution, expressing electrolyte water content, is compared with that obtained by means of a complex numerical approach, used to solve the same mathematical problem. The achieved results show that the mean error is below 5% for electrodes water content values ranging from 2 to 15 (given as boundary conditions), and it does not overcome 0.26% for electrodes water content above 5. These results prove the capability of the solution to correctly model electrolyte water content at any operating condition, aiming at embodiment into more complex frameworks (e.g., cell or stack models), related to fuel cell simulation, monitoring, control, diagnosis and prognosis.

Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology - KTH, Roslagstullsbacken 21, S-10691 Stockholm (Sweden)]. E-mail: alby@anl.gov

2007-01-15

We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in {sup 235}U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides {sup 240}Pu, {sup 238}U and {sup 232}Th and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for {sup 240}Pu, {sup 238}U and {sup 232}Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 {mu}m and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core.

Nuclear power plant types and the management of plutonium and minor actinides - in search of fuel cycle flexibility

International Nuclear Information System (INIS)

Thomas, J.B.

2002-01-01

Transuranics management concerns all NPP types, because of the specifications for sustainable development. Multiple recycling is mandatory. Neutronic abundance can be obtained in fast spectrum, or by adding external neutrons or (temporarily) with additional 235 U. The LWRs can control the plutonium inventory and significantly reduce the amount of transuranics transferred to the geological repository, thanks to the use of innovative nuclear fuel in a limited part of the NPP fleet. HTR adapted to transuranics burning can help. In the future, in addition to the liquid metal FBR, a strategy based on a gas cooled technological line and advanced fuel opens a second path towards fast spectra. Strategies for defining the optimal mix of reactor types in the nuclear fleet at a given time and demonstrating the fuel cycle flexibility are under study. (author)

Dynamic buckling and nonlinear response of FBR main vessels under earthquake loading

International Nuclear Information System (INIS)

Hagiwara, Yutaka; Kawamoto, Yoji; Nakagawa, Masaki; Akiyama, Hiroshi.

1991-01-01

Pseudo-dynamic tests of cylindrical shells under high temperature were performed in order to study elasto-plastic shear-bending buckling and the nonlinear response of FBR main vessels under earthquake loading. The test results showed a response reduction effect due to pre-buckling plasticity, and a large seismic margin due to post-buckling energy absorption of the cylinders. A simple expression of the response reduction effect was proposed, as a contribution to the safe and effective seismic design of FBRs. Two methods for seismic margin evaluation were also proposed, and it was shown that appropriate seismic margins can be ensured, when the response reduction effect is incorporated into the seismic design. (author)

SRL online Analytical Development

International Nuclear Information System (INIS)

Jenkins, C.W.

1991-01-01

The Savannah River Site is operated by the Westinghouse Savannah River Co. for the Department of Energy to produce special nuclear materials for defense. R ampersand D support for site programs is provided by the Savannah River Laboratory, which I represent. The site is known primarily for its nuclear reactors, but actually three fourths of the efforts at the site are devoted to fuel/target fabrication, fuel/target reprocessing, and waste management. All of these operations rely heavily on chemical processes. The site is therefore a large chemical plant. There are then many potential applications for process analytical chemistry at SRS. The Savannah River Laboratory (SRL) has an Analytical Development Section of roughly 65 personnel that perform analyses for R ampersand D efforts at the lab, act as backup to the site Analytical Laboratories Department and develop analytical methods and instruments. I manage a subgroup of the Analytical Development Section called the Process Control ampersand Analyzer Development Group. The Prime mission of this group is to develop online/at-line analytical systems for site applications

Development of numerical and analytical methodology for stress analysis in guide tubes of fuel elements

International Nuclear Information System (INIS)

Carrilho, Leo A.; Dotto, Rosvita M.; Gouvea, Jayme P. de

2000-01-01

The stresses in the components of fuel elements in operation have been calculated by Industrias Nucleares do Brasil - INB, using programmes specifically developed for this are. However, worldwide useful software as Excel and ANSYS have resources that make them an alternative with advantages for those computing. In this context, the stress and displacements were calculated in the guide thimbles of a fuel element in normal operation in the reactor under static loads, through analytic and numeric models, which results are comparable to that obtained with the actual INB's methodology. The discussion of the results exposes the peculiarity of a pick of compression stress in a segment of the guide thimble which is accentuated during low power operations. Suggestions for the relief of these high stresses are proposed for future studies. (author)

Fabrication of 0.5-inch diameter FBR mixed oxide fuel pellets

International Nuclear Information System (INIS)

Rasmussen, D.E.; Benecke, M.W.; McCord, R.B.

1979-01-01

Large diameter (0.535 inch) mixed oxide fuel pellets for Fast Breeder Reactor application were successfully fabricated by the cold-press-and-sinter technique. Enriched UO 2 , PuO 2 -UO 2 , and PuO 2 -ThO 2 compositions were fabricated into nominally 90% theoretical density pellets for the UO 2 and PuO 2 -UO 2 compositions, and 88% and 93% T.D. for the PuO 2 -ThO 2 compositions. Some processing adjustments were required to achieve satisfactory pellet quality and density. Furnace heating rate was reduced from 200 to 50 0 C/h for the organic binder burnout cycle for the large, 0.535-inch diameter pellets to eliminate pellet cracking during sintering. Additional preslugging steps and die wall lubrication during pressing were used to eliminate pressing cracks in the PuO 2 -ThO 2 pellets

A real-time data acquisition and processing system for the analytical laboratory automation of a HTR spent fuel reprocessing facility

International Nuclear Information System (INIS)

Watzlawik, K.H.

1979-12-01

A real-time data acquisition and processing system for the analytical laboratory of an experimental HTR spent fuel reprocessing facility is presented. The on-line open-loop system combines in-line and off-line analytical measurement procedures including data acquisition and evaluation as well as analytical laboratory organisation under the control of a computer-supported laboratory automation system. In-line measurements are performed for density, volume and temperature in process tanks and registration of samples for off-line measurements. Off-line computer-coupled experiments are potentiometric titration, gas chromatography and X-ray fluorescence analysis. Organisational sections like sample registration, magazining, distribution and identification, multiple data assignment and especially calibrations of analytical devices are performed by the data processing system. (orig.) [de

Dynamic analytical and experimental research of shock absorber to safeguard the nuclear fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Dundulis, Gintautas, E-mail: gintas@mail.lei.lt [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Grybenas, Albertas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Karalevicius, Renatas [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Makarevicius, Vidas [Laboratory of Materials Research and Testing, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); Rimkevicius, Sigitas; Uspuras, Eugenijus [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas (Lithuania)

2013-07-15

Highlights: • Plastical deformation of the shock absorber. • Dynamic testing of the scaled shock absorber. • Dynamic simulation of the shock absorber using finite element method. • Strain-rate evaluation in dynamic analysis. • Variation of displacement, acceleration and velocity during dynamic impact. -- Abstract: The Ignalina Nuclear Power Plant (NPP) has two RBMK-1500 graphite moderated boiling water multi-channel reactors. The Ignalina NPP Unit 1 was shut down at the end of 2004 while Unit 2 has been in operation for over 5 years. After shutdown at the Unit 1 remained spent fuel assemblies with low burn-up depth. In order to reuse these assemblies in the reactor of Unit 2 a special set of equipment was developed. One of the most important items of this set is a container, which is used for the transportation of spent fuel assemblies between the reactors of Unit 1 and Unit 2. A special shock absorber was designed to avoid failure of fuel assemblies in case of hypothetical spent fuel assemblies drop accident during uploading/unloading of spent fuel assemblies to/from container. This shock absorber was examined using scaled experiments and finite element analysis. Static and dynamic investigations of the shock absorber were performed for the estimation and optimization of its geometrical parameters. The objective of this work is the estimation whether the proposed design of shock absorber can fulfil the stopping function of the spent fuel assemblies and is capable to withstand the dynamics load. Experimental testing of scaled shock absorber models and dynamic analytical investigations using the finite element code ABAQUS/Explicit were performed. The simulation model was verified by comparing the experimental and simulation results and it was concluded that the shock absorber is capable to withstand the dynamic load, i.e. successful force suppression function in case of accident.

An overview of analytical activities of control laboratory in NFC

International Nuclear Information System (INIS)

Balaji Rao, Y.; Subba Rao, Y.; Saibaba, N.

2015-01-01

As per the mandate of Department of Atomic Energy (DAE), Nuclear Fuel Complex (NFC) was established in 1971 for manufacturing Fuel Sub-assemblies for both PHWRs and BWRs operating in India on industrial scale. Control Laboratory (C.Lab) was envisaged as a centralized analytical facility to achieve the objectives of NFC on the similar lines of its predecessor, Analytical Chemistry Division at BARC. With highest ever production of 1200 MT of PHWR Fuel and 16 lakhs PHWR Fuel Tubes achieved during production year of 2014-15 and with increase in demand further for fuel requirements, NFC has got demanding situation in next year and accordingly, C. Lab has also geared up to meet the challenging demands of all the production plant. The average annual analytical load comes around 5 Lakhs estimations and to manage such a massive analytical load a proper synergy between good chemistry, process conditions and analytical methods is a necessity and laboratory is able to meet this important requirement consistently

Alternative fuels for the French fast breeder reactors programme

International Nuclear Information System (INIS)

Bailly, H.; Bernard, H.; Mansard, B.

1988-01-01

At the present time, due to the very competitive cost per kWh produced in France by the PWRs, it appears clear that, despite the improved use of uranium by FBRs, they will only be developed if the cost of the fuel cycle is sufficiently lower than that of the PWRs to compensate for the additional investment. The current economic programme has fixed the following fuel related objectives: - burn-up as high possible, the value of 150 000 MWd/t being considered as a minimum, and not a final target to be achieved, - extension of the duration of reactor operation cycles, leading to high in-pile times for fuel. Reaching the latter objective depends on obtaining high internal breeding gain performances, so that the total reactivity drop related to fuel impoverishment can be minimized. In this respect, a large diameter oxide fuel and/or an axial heterogeneous core concept can be envisaged. Dense fuels could form another solution. The feasibility of the fabrication of carbide and nitride fuels has been demonstrated in several countries and there is currently convergence towards a single type of process based on a carbothermic reaction. The optimization of fabrication procedures for these fuels must be continued to satisfy economic requirements and to obtain a fabrication cost of the same order or magnitude as that of oxide, although higher. If this target is achieved, fabrication will not be the major criterion for the selection of the FBR fuel, which will then be a function of the cost of reprocessing, performances under irradiation and reactor operating requirements

A reduced chemical kinetic model for the analytical investigations on the oxidation kinetics and performance characteristics of diesel fuel

International Nuclear Information System (INIS)

Selvaraj, N.; Manoj Kumar, C.V.; Babu, M.S.

2010-01-01

A detailed study of the combustion of diesel fuel has been conducted analytically using a kinetic scheme with 767 elementary reactions and 158 species. A program has been developed in MATLAB for the analysis of ignition delay, performance, soot formation and emission characteristics of diesel fuel. Nitrogen is considered as the diluent and its percentage is assumed as 79%. The criteria used for the determination of ignition delay time are based on OH concentration to reach a value of 1x10 -9 . A brief review of diesel combustion and soot formation is given. (author)

Aluminum and aluminum/silicon coatings on ferritic steels by CVD-FBR technology

International Nuclear Information System (INIS)

Perez, F.J.; Hierro, M.P.; Trilleros, J.A.; Carpintero, M.C.; Sanchez, L.; Bolivar, F.J.

2006-01-01

The use of chemical vapor deposition by fluidized bed reactors (CVD-FBR) offers some advantages in comparison to other coating techniques such as pack cementation, because it allows coating deposition at lower temperatures than pack cementation and at atmospheric pressure without affecting the mechanical properties of material due to heat treatments of the bulk during coating process. Aluminum and aluminum/silicon coatings have been obtained on two different ferritics steels (P-91 and P-92). The coatings were analyzed using several techniques like SEM/EDX and XRD. The results indicated that both coatings were form by Fe 2 Al 5 intermetallic compound, and in the co-deposition the Si was incorporated to the Fe 2 Al 5 structure in small amounts

Air pollution impacts from logistics related to forest biomass to energy chain

Energy Technology Data Exchange (ETDEWEB)

Monteiro, C.; Tarelho, L.; Lopes, M.; Monteiro, A.; Cascao, P.; Miranda, A.M. [CESAM and Dept. of Environment and Planning, Univ. of Aveiro, Aveiro (Portugal)], e-mail: anacristina@ua.pt

2012-11-01

In recent years, pressures on global environment and energy security have led to an increasing demand on renewable energy sources, and diversification of world's energy supply. Among these resources the forest biomass could exert an important role, since it is considered a renewable and CO{sub 2} neutral energy resource, and can potentially provide energy for heat, power and transport fuels. In this study were presented the results of the amounts of forest biomass residues (FBR) available in Portugal, taking into account some conditioning related with land characteristics (e.g. slope). Comparing the FBR consumption for industrial thermal power plants it is possible to verify that the FBR available (1.91x10{sup 6} ton (dry) year{sup -}1) in Portugal is enough to address the needs of industrial plants, but if the planned plants come into operation, the FBR available is no longer sufficient. The operations associated with the FBR harvesting were described, emphasizing the transport between the production locations and the industrial thermal plants. By applying a TRaffic Emission Model (TREM), it was estimated the fuel consumption and related gaseous emissions (CO, CO{sub 2}, PM, NO{sub x}, SO{sub 2}, VOC, CH{sub 4}, NH{sub 3} and N{sub 2}O) associated with the transport of the FBR.

Analytic models for fuel pin transient performance

International Nuclear Information System (INIS)

Bard, F.E.; Fox, G.L.; Washburn, D.F.; Hanson, J.E.

1976-09-01

HEDL's ability to analyze various mechanisms that operate within a fuel pin has progressed substantially through development of codes such as PECTCLAD, which solves cladding response, and DSTRESS, which solves fuel response. The PECTCLAD results show good correlation with a variety of mechanical tests on cladding material and also demonstrate the significance of cladding strength when applying the life fraction rule. The DSTRESS results have shown that fuel deforms sufficiently during overpower transient tests that available volumes are filled, whether in the form of a central cavity or start-up cracks

Integrated thermal analysis of top-shield and reactor vault of Indian FBR-600

International Nuclear Information System (INIS)

Rajendrakumar, M.; Velusamy, K.; Selvaraj, P.

2015-01-01

The design for next generation fast breeder reactors (FBR-600) has been commenced with enhanced safety and improved economy as the main targets. The Top Shield (TS) of Prototype Fast Breeder Reactor (PFBR) is a box type structure consisting of Roof Slab (RS), Small Rotatable Plug (SRP), and Large Rotatable Plug (LRP). The large box type structure with many penetrations posed difficulties during manufacturing. Because of the required high load carrying capabilities, a dome shaped thick plate roof slab is conceived for FBR-600. Main Vessel (MV) which holds the primary sodium and associated components is welded to the RS through a triple joint. Reactor vault (RV) is a thick concrete structure which supports MV and Safety Vessel (SV). The temperature of RV concrete has to be less than 338 K (65°C) under normal operating heat loads (full and part load conditions) and less than 363 K (90°C) under Safety Grade Decay Heat Removal (SGDHR) conditions with one cooling loop in service. The temperature in the component penetrations of the RS should be greater than 120°C to avoid sodium aerosol deposition. Similarly, the temperature of the LRP and SRP has to be ∼120°C to protect the elastomeric seals provided to these structures. Further, the heat load to RV transferred by direct conduction by roof slab support has to be minimum. To meet these conflicting thermal requirements, detailed multi-physics CFD calculations have been performed to finalize, (i) the insulation requirements on the top of roof slab, (ii) number and position of reflective insulation plates below the bottom plate of roof slab/rotating plugs, (iii) air flow rate for various zones of the top shield and (iv) water flow rate and pitch of water cooling pipes for the reactor vault. (author)

Development and evaluation of analytical techniques for total chlorine in burner fuels

International Nuclear Information System (INIS)

Gaskill, A. Jr.; Estes, E.D.; Hardison, D.L.; Friedman, P.H.

1987-01-01

A current EPA regulation prohibits the sale for burning in non-industrial boilers of used oils and oil fuels contaminated above specified levels with certain metals and total chlorine. When burned as fuel in a small boiler, the contaminants may be emitted to the ambient air at hazardous levels. This regulation establishes a rebuttable presumption that used oil containing more than 1,000 ppm total chlorine has been mixed with halogenated solvents and is a hazardous waste. Rebutting the presumption requires the seller of the oil to prove that this chlorine is not due to halogenated solvents or other hazardous halogenated organics. If the rebuttal is successful, the oil can be sold as fuel up to a level of 4000 ppm total chlorine. Analytical techniques for determination of total chlorine were evaluated or developed to provide regulatory agencies and the regulated community with appropriate chlorine test methods. The techniques evaluated included chemical titrations following oxygen bomb combustion, disposable field test kits, instrumental microcoulometry, and x-ray fluorescence spectrometry. These candidate techniques were subjected to interlaboratory testing to estimate their precision, accuracy, sensitivity, and susceptibility to matrix effects. Information on ease of use and analysis costs was also collected. Based on this pilot study, test methods will be written for the most promising techniques and subjected to a formal collaborative study to generate precision and accuracy data for each method. These methods are to be proposed in the Federal Register as mandatory for compliance with the existing used oil regulation

Effects of fast breeders' characteristics on consumptions and expenditures related to electric power generation. An assessment for the European Community

International Nuclear Information System (INIS)

Graziani, G.; Zanantoni, C.

1976-12-01

The effects of the physical characteristics of the FBR (breeding ratio, inventory) and of the associated fuel cycle (out-of-pile time) on the results of strategy calculations performed for an electric power system are discussed. Attention is focused on the penetration of FBR, on the demand of uranium and separative work and on the economical performance of the breeders. It is pointed out that uncertainties about the out-of-pile time are as important as those concerning FBR breeding ratio and inventory

Fuel pins irradiation: experimental devices and analytical behaviour

International Nuclear Information System (INIS)

Lemaignan, C.

1996-01-01

In this text we present the general characteristics of adapted irradiation loops in research reactors and the main results that we can expected with these loops in the behaviour field of PWR and LMFBR fuels( fuel densification, fuel cladding interactions, fission products release, reactor accidents)

Survey of evaluation methods for thermal striping in FBR structures

International Nuclear Information System (INIS)

Miura, Naoki; Nitta, Akito; Take, Kohji

1988-01-01

In the upper core structures or the sodium mixing tee of Fast Breeder Reactors, sodium mixing streams which are at different temperatures produce rapid temperature fluctuations, namely 'thermal striping', upon component surfaces, and it is apprehended that the high-cycle thermal fatigue causes the crack initiation and propagation. The thermal striping is one of the factors which is considered in FBR component design, however, the standard evaluation method has not built up yet because of the intricacy of that mechanism, the difficulty of an actual proof, the lack of data, and so on. In this report, it is intended to survey of the datails and the present situation of the evaluation method of crack initiation and propagation due to thermal striping, and study the appropriate method which will be made use of the rationalization of design. So it is ascertained that the method which use a quantitative prediction of crack propagation is optimum to evaluate the thermal striping phenomenon. (author)

Synergistic fuel cycles of the future

International Nuclear Information System (INIS)

Meneley, D.A.; Dastur, A.R.

1997-01-01

Good neutron economy is the basis of the fuel cycle flexibility in the CANDU reactor. This paper describes the fuel cycle options available to the CANDU owner with special emphasis on resource conservation and waste management. CANDU fuel cycles with low initial fissile content operate with relatively high conversion ratio. The natural uranium cycle provides over 55 % of energy from the plutonium that is created during fuel life. Resource utilization is over 7 MWd/kg NU. This can be improved by slight enrichment (between 0.9 and 1.2 wt % U235) of the fuel. Resource utilization increases to 11 MWd/kg NU with the Slightly Enriched Uranium cycle. Thorium based cycles in CANDU operate at near-breeder efficiency. Obey provide attractive options when used with natural uranium or separated (reactor grade and weapons grade) plutonium as driver fuels. In the latter case, the energy from the U233 plus the initial plutonium content amounts to 3.4 GW(th).d/kg Pu-fissile. The same utilization is expected from the use of FBR plutonium in a CANDU thorium cycle. Extension of natural resource is achieved by the use of spent fuels in CANDU. The LWR/CANDU Tandem cycle leads to an additional 77 % of energy through the use of reprocessed LWR fuel (which has a fissile content of 1.6 wt %) in CANDU. Dry reprocessing of LWR fuel with the OREOX process (a more safeguardable alternative to the PUREX process) provides an additional 50 % energy. Uranium recovered (RU) from separation of plutonium contained in spent LWR fuel provides an additional 15 MWd/kg RU. CANDU's low fissile requirement provides the possibility, through the use of non-fertile targets, of extracting energy from the minor actinides contained in spent fuel. In addition to the resource utilization advantage described above, there is a corresponding reduction in waste arisings with such cycles. This is especially significant when separated plutonium is available as a fissile resource. (author)

Analytical approximations for the long-term decay behavior of spent fuel and high-level waste

International Nuclear Information System (INIS)

Malbrain, C.M.; Deutch, J.M.; Lester, R.K.

1982-01-01

Simple analytical approximations are presented that describe the radioactivity and radiogenic decay heat behavior of high-level wastes (HLWs) from various nuclear fuel cycles during the first 100,000 years of waste life. The correlations are based on detailed computations of HLW properties carried out with the isotope generation and depletion code ORIGEN 2. The ambiguities encountered in using simple comparisons of the hazards posed by HLWs and naturally occurring mineral deposits to establish the longevity requirements for geologic waste disposal schemes are discussed

Creep-fatigue life property of FBR high-temperature structural materials under tension-torsion loading and life evaluation method

International Nuclear Information System (INIS)

Ogata, Takashi; Nitta, Akito

1994-01-01

Creep-fatigue damage in high temperature structural components in a FBR progress under multiaxial stress condition depending on their operating conditions and configuration. Therefore, multiaxial stress effects on creep-fatigue damage evolution must be clarified to make precise creep-fatigue damage evaluation of these components. In this study, creep-fatigue tests in FBR high temperature materials such as SUS304, 316FR stainless steels and a modified 9Cr steel were conducted under biaxial stress subjecting tension-compression and torsion loading, in order to examine biaxial stress effects on failure mechanism and life property, and to discuss creep-fatigue life evaluation methods under biaxial stress. Main results obtained in this study are summarized as follows: 1. The main cracks under cyclic torsion loading propagated by shear mode in three materials. But intergranular failure was occurred in SUS304 and 316FR, and transgranular failure was observed in Mod.9Cr steel. 2. Nonlinear damage accumulation model proposed based on uniaxial creep-fatigue test results was extended to apply for creep-fatigue damage evaluation under biaxial stress state by considering the biaxial stress effects on fatigue and creep damage evolution. 3. It was confirmed that creep-fatigue life under biaxial stress could be predicted by the extended evaluation method with higher accuracy than existing methods. (author)

Determination of trace impurities in uranium-transition metal alloy fuels by ICP-MS using extended common analyte internal standardization (ECAIS) technique

International Nuclear Information System (INIS)

Saha, Abhijit; Deb, S.B.; Nagar, B.K.; Saxena, M.K.

2015-01-01

An analytical methodology was developed for the determination of eight trace impurities viz, Al, B, Cd, Co, Cu, Mg, Mn and Ni in three different uranium-transition metal alloy fuels (U-Me; Me = Ti, Zr and Mo) employing inductively coupled plasma mass spectrometry (ICP-MS). The well known common analyte internal standardization (CAIS) chemometric technique was modified and then employed to minimize and account for the matrix effect on analyte intensity. Standard addition of analytes to the pure synthetic U-Me sample solutions and subsequently their ≥ 94% recovery by the ICP-MS measurement validates the proposed methodology. One real sample of each of these alloys was analyzed by the developed analytical methodology and the %RSD observed was in the range of 5-8%. The method detection limits were found to be within 4-10 μg L -1 . (author)

Outline of environmental impact of waste management

International Nuclear Information System (INIS)

1979-09-01

This document presents background information on the environmental impacts from the management and disposal of radioactive waste for seven reference fuel cycles selected by INFCE Working Group 7, but excluding the health and safety impact on man. The main factors considered were: use of natural resources, land, water, energy, labour and materials; effects of chemical and thermal effluents; effects of meteorology, hydrology and natural hazards; and social effects. The environmental impacts are generally largest for the once-through fuel cycles and smallest for the FBR and HWR U/Th cycles, due to the impacts being correlated to uranium requirements. The main impact is the use of land which varies from 0.1 - 1.6 ha/GWa with the FBR strategy requiring the smallest use of land and the LWR once-through strategy the largest. The land use for mill tailings is, except for the FBR and U/Th cycles, dominant compared to the land use for the rest of the fuel cycle

Technology readiness of partitioning and transmutation toward closed fuel cycle in Japan

International Nuclear Information System (INIS)

Ikeda, Kazumi; Kurata, Masaki; Morita, Yasuji; Tsujimoto, Kazufumi; Minato, Kazuo; Koyama, Shin-ichi

2011-01-01

This paper treats technology readiness level (TRL) assessment of Partitioning and Transmutation (P-T) toward closed fuel cycle in JAPAN. The purpose is providing clarified information related to the current maturity of the partitioning and transmutation technologies by applying the methodology of TRL, parallel to attempting to establish common indications among relating technology area. The methodology should be one of useful communication tools between specialists and management level, and also among countries interested in the P-T technologies. The generic TRL in this study is based on the GNEP (Global Nuclear Energy Partnership)'s definition: TRL 3 shows the status that critical function is proved and elemental technologies are identified, TRL 4 represents that relating technologies are validated at bench scale in laboratory environment, and TRL 5 achieves the completion of development related to the subsystem and elemental technologies. Detailed indications are established through discussion of the relating specialists. Reviewed technological area includes P-T and minor actinide (MA) cycle: Fast Breeder Reactor (FBR) and Accelerator driven system (ADS) for MA transmutation, partitioning processes, and MA-bearing fuels. The assessments reveal that TRL spreads around TRL 3 to TRL 4 because each system requires more the development of elemental technologies. Transmutation core of FBR is assessed to be TRL 4 in that MA bearing integral test is required additionally, and ADS becomes TRL 3 because the elemental technologies were identified and the requirements were specified. Consequently, the common key issue is how the nuclear calculation methodology will be validated for MA-bearing-fuelled core, since several percentages of MA changes the void reactivity and the Doppler Effect significantly, which are inherently important in reactor safety. It should be that critical experiments with several kg of americium or more are difficult in the existing experimental

The secondary stress analyses in the fuel pin cladding due to the swelling gradient across the wall thickness

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ukai, Shigeharu

2002-01-01

Irradiation deformation analyses of FBR fuel cladding were made by using the finite element method. In these analyses the history of the stress occurred in the cladding was evaluated paying attention to the secondary stress induced by the swelling difference across the wall thickness. It was revealed that the difference of the swelling incubation dose in the direction of the thickness and the irradiation creep deformation play an important role in the history of the secondary stress. The effect of the stress-enhanced swelling was also analyzed in this study

Hydrogen transfer preventive device in FBR power plant

International Nuclear Information System (INIS)

Hoshi, Yuichi.

1987-01-01

Purpose: To prevent transfer of hydrogen, etc. in FBR power plant. Constitution: Since H 2 permeates heat conduction pipes in a steam generator, it is necessary to eliminate all of permeation hydrogen, etc. by primary cold traps particularly in the case of saving the intermediate heat exchange. In view of the above, the heat conduction pipes of the steam generator are constituted as a double pipe structure and helium gases are recycled through the gaps thereof and hydrogen traps are disposed to the recycling path. H 2 released into water flowing through the inside of the inner pipe is permeated through the inner pipe and leached into the gap, but the leached H 2 is carried by the helium recycling stream to the hydrogen trap and then the H 2 stream removed with H 2 is returned to the gaps. In this way, the capacity of the primary cold traps disposed in the liquid sodium recycling circuit can be reduced remarkably and the capacity of the purifying device, if an intermediate heat exchanger is disposed, is also reduced to decrease the plant cost. Further, diffusion of deleterious gases from the primary to the secondary circuits can be prevented as well. (Kamimura, M.)

Characteristics of polyethylene and zirconium-hydride moderator for the NSRR tests

International Nuclear Information System (INIS)

Nakamura, Takehiko; Yamazaki, Toshi; Sobajima, Makoto.

1994-03-01

Pulse irradiation tests of FBR fuels under the sodium cooling conditions are planned for the phase III program in the NSRR (Nuclear Safety Research Reactor), following the phase I and II programs of the LWR fuel tests under the simulated RIA (Reactivity Initiated Accident) conditions. A proto-type irradiation capsule for the FBR fuel rod tests and a sodium loop to purify and to charge sodium into the capsule are under construction for the tests. In the NSRR tests, neutron moderator is needed to thermalize neutrons from the driver core and to subject transient energy high enough to cause the test fuel failure. The light water has been used for the NSRR LWR fuel tests as the coolant/moderator material. Polyethylene and zirconium-hydride are candidates of the moderator for the FBR fuel tests. The capability of the moderators are investigated in the pulse irradiation tests in the NSRR. Both of the moderators indicated good capability of realizing high thermal neutron flux to subject energy depositions comparable to the light water or higher. Estimations by the SRAC code system indicated reasonable good agreement with the test results. In addition, heating tests of the moderators did not cause gas decomposition nor dissociation, indicating that the moderators are operative at temperatures up to 300degC. (author)

Modeling of the anode side of a direct methanol fuel cell with analytical solutions

International Nuclear Information System (INIS)

Mosquera, Martin A.; Lizcano-Valbuena, William H.

2009-01-01

In this work, analytical solutions were derived (for any methanol oxidation reaction order) for the profiles of methanol concentration and proton current density, by assuming diffusion mass transport mechanism, Tafel kinetics, and fast proton transport in the anodic catalyst layer of a direct methanol fuel cell. An expression for the Thiele modulus that allows to express the anodic overpotential as a function of the cell current and kinetic and mass transfer parameters was obtained. For high cell current densities, it was found that the Thiele modulus (φ 2 ) varies quadratically with cell current density; yielding a simple correlation between anodic overpotential and cell current density. Analytical solutions were derived for the profiles of both local methanol concentration in the catalyst layer and local anodic current density in the catalyst layer. Under the assumptions of the model presented here, in general, the local methanol concentration in the catalyst layer cannot be expressed as an explicit function of the position in the layer. In spite of this, the equations presented here for the anodic overpotential allow the derivation of new semi-empirical equations

Analytical Evaluation to Determine Selected PAHs by HPLC in a Type 2 Fuel; Evaluacion Analitica de 4 Metodos de Determinacion de PAHs medianteHPLC en un Fuel de Tipo II

Energy Technology Data Exchange (ETDEWEB)

Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Escolano Segovia, O.; Garcia Frutos, F. J.

2009-05-21

An evaluation of analytical parameters to determine selected PAHs in a fuel oil type II by HPLC coupled to fluorescence and diode detectors is presented. The study was focused on four conventional treatments of these kinds of oil samples and the main objective was giving a measure of confidence level of PAH results in the fuel oil. This study was performed in the frame of the project Assessment of natural attenuation of PAHs in agricultural soil contaminated with fuel from an accidental spill (Spanish National Plain I+D+I, CTM2007-64537). This paper is presented as follows: Analysis of reference material 1582 (NIST) by using the four kinds of sample treatments of interest. Application of variance analysis to compare results obtained from type II fuel by using each sample treatment and chromatographic detector. Finally, a statistic calculation was performed to measure uncertainty components in chromatographic analysis. (Author)

Mass transfer of steels for FBR in sodium loop

International Nuclear Information System (INIS)

Susukida, Hiroshi; Yonezawa, Toshio; Ueda, Mitsuo; Imazu, Takayuki; Kiyokawa, Teruyuki.

1976-06-01

In order to grasp quantitatively the corrosion and mass transfer of steels for FBR in sodium loop and to establish their allowable stress value and corrosion rate, a special sodium loop for material testing was designed and fabricated and the steels were given 3010 hours exposing test in the sodium loop. This paper gives the outline of the sodium loop and the results of the test. (1) Carburization and a slight increase in weight were observed in the specimens of type 304 stainless steel exposed in the sodium loop for 3010 hours, while decarburization was observed in the specimens of 2 1/4 Cr-1 Mo steel. It is considered that these phenomena were caused by the downstream factor of the sodium loop. (2) A remarkable decrease of Charpy absorbed energy was observed in the specimens of type 304 stainless steel exposed in the sodium loop. It is considered that this resulted from the weakening of the grain boundary due to heat history and mass transfer. (3) The specimens exposed in the sodium loop must be washed by ultrasonic waves in a water bath after washing in alcohol. (auth.)

The Analytical Repository Source-Term (AREST) model: Analysis of spent fuel as a nuclear waste form

International Nuclear Information System (INIS)

Apted, M.J.; Liebetrau, A.M.; Engel, D.W.

1989-02-01

The purpose of this report is to assess the performance of spent fuel as a final waste form. The release of radionuclides from spent nuclear fuel has been simulated for the three repository sites that were nominated for site characterization in accordance with the Nuclear Waste Policy Act of 1982. The simulation is based on waste package designs that were presented in the environmental assessments prepared for each site. Five distinct distributions for containment failure have been considered, and the release for nuclides from the UO 2 matrix, gap (including grain boundary), crud/surface layer, and cladding has been calculated with the Analytic Repository Source-Term (AREST) code. Separate scenarios involving incongruent and congruent release from the UO 2 matrix have also been examined using the AREST code. Congruent release is defined here as the condition in which the relative mass release rates of a given nuclide and uranium from the UO 2 matrix are equal to their mass ratios in the matrix. Incongruent release refers to release of a given nuclide from the UO 2 matrix controlled by its own solubility-limiting solid phase. Release of nuclides from other sources within the spent fuel (e.g., cladding, fuel/cladding gap) is evaluated separately from either incongruent or congruent matrix release. 51 refs., 200 figs., 9 tabs

Biodiesel Analytical Methods: August 2002--January 2004

Energy Technology Data Exchange (ETDEWEB)

Van Gerpen, J.; Shanks, B.; Pruszko, R.; Clements, D.; Knothe, G.

2004-07-01

Biodiesel is an alternative fuel for diesel engines that is receiving great attention worldwide. The material contained in this book is intended to provide the reader with information about biodiesel engines and fuels, analytical methods used to measure fuel properties, and specifications for biodiesel quality control.

Temperature monitoring and leak detection in sodium circuits of FBR using Raman distributed fiber optic sensor

International Nuclear Information System (INIS)

Kasinathan, M.; Murali, N.; Sosamma, S.; Babu Rao, C.; Kumar, Anish; Purnachandra Rao, B.; Jayakumar, T.

2013-01-01

This paper discusses the fiber optic temperature sensor based leak detection in the coolant circuits of fast breeder reactor. These sensors measure the temperature based on spontaneous Raman scattering principle and is not influenced by the electromagnetic interference. Various experiments were conducted to evaluate the performance of the fiber optic sensor based leak detection using Raman distributed Temperature Sensor (RDTS). This paper also deals with the details of fiber optic sensor type leak detector layout for the coolant circuit of FBR, performance requirement of leak detection system, description of the test facility, experimental procedure and test results of various experiments conducted. (author)

A review of fast reactor program in Japan

International Nuclear Information System (INIS)

1992-01-01

In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

A review of fast reactor program in Japan

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods

International Nuclear Information System (INIS)

Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C.

1979-01-01

Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility (STF). One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. (orig.)

D.E.M.T. Experimental and analytical studies on seismic isolation

International Nuclear Information System (INIS)

Gantenbein, F.; Buland, P.

1989-01-01

The various studies which have been performed in C.E.A./D.E.M.T. will be reviewed in the paper. They are experimental or theoretical and related to the overall behavior of isolated structures. Among the experimental work one can notice: - the seismic tests on a shaking table of a concrete cylinder isolated by sliding neoprene pads, - the vibrational tests on the reaction mass of TAMARIS seismic facility. The analytical work consists of dynamic calculation method development: - for the soil structure interaction in case of pads interposed between an upper raft and pedestals; - for the time history calculation of sliding structures; - for fluid structure interaction (coupling of fluid and structure motion or sloshing modes). Finally comments will be given on the seismic isolation consequencies for the analysis of F.B.R. vessels: the modes can no more be considered independent (SRSS method leads to important errors), the sloshing increases

Simulation of radioactive corrosion product in primary cooling system of Japanese sodium-cooled fast breeder reactor

International Nuclear Information System (INIS)

Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

2012-01-01

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

Development of FBR integrity system code. Basic concept

International Nuclear Information System (INIS)

Asayama, Tai

2001-05-01

For fast breeder reactors to be commercialized, they must be more reliable, safer, and at the same, economically competitive with future light water reactors. Innovation of elevated temperature structural design standard is necessary to achieve this goal. The most powerful way is to enlarge the scope of structural integrity code to cover items other than design evaluation that has been addressed in existing codes. Items that must be newly covered are prerequisites of design, fabrication, examination, operation and maintenance, etc. This allows designers to choose the most economical combination of design variations to achieve specific reliability that is needed for a particular component. Designing components by this concept, a cost-minimum design of a whole plant can be realized. By determining the reliability that must be achieved for a component by risk technologies, further economical improvement can be expected by avoiding excessive quality. Recognizing the necessity for the codes based on the new concept, the development of 'FBR integrity system code' began in 2000. Research and development will last 10 years. For this development, the basic logistics and system as well as technologies that materialize the concept are necessary. Original logistics and system must be developed, because no existing researches are available in and out of Japan. This reports presents the results of the work done in the first year regarding the basic idea, methodology, and structure of the code. (author)

Fast reactors: the future of nuclear energy

International Nuclear Information System (INIS)

Carvalho, H.G. de.

1988-08-01

The main problems to be solved for FBR type reactors become viable economically, presenting the research programs of Europe, United States of America, Japan and Brazil are described. The cooperations between interested countries for improving FBR type reactors, and the financial and human resources necessaries for the development of programs, are evaluated. The fuel cycle is also analysed. (M.C.K.) [pt

Application of non-reductive partial partitioning in FBR fuel reprocessing: a simulation study

International Nuclear Information System (INIS)

Shekhar Kumar; Koganti, S.B.

2002-06-01

The observed performance of conventional partitioning contactor in the Purex process in the Purex process is seldom satisfactory due to over-consumption of reductant and poor U-Pu decontamination factors. Contemporary trends indicate gradually move-over to MOX fuels for FBRs. In this scenario, it is not necessary to separate uranium and plutonium completely. By controlling the acid concentration and flow rates, it is possible to selectively strip essentially all the plutonium and part of the uranium in the aqueous stream. Therefore, a mixed product enriched in plutonium is obtained which can be precipitated and denitrated. Alternatively direct denitration by microwave heating can be used. The idea is particularly attractive for the flowsheet meant for the first core of FBTR where Pu/(U+Pu) ratio of 0.7 (in the discharged fuel) is diluted by addition of uranium to 0.3. By partial partitioning in the 2B contactor, a product enriched in plutonium (Pu/(U+Pu) ratio ∼0.6) can be obtained. After a minor uranium addition, this product will be suitable for the direct fabrication of II core of FBTR where a Pu/(U+Pu) ratio of 0.55 will be required. In a similar fashion, the enriched product can be used for multiple core zones of proposed PFBR by selective additions of uranium for each zone. To explore the feasibility of such partial partitioning step, an exhaustive simulation study was made using the in-house developed computer code SIMPSEX. 1300 simulations runs were completed for different combinations of parameters and results were analyzed. In this report, the results of this study and a possible flowsheet step have been discussed. Simultaneous variation in the flow rates has been considered and safe operating limits for the partial partitioning step have been established. (author)

Promotion of research and development for fast breeder reactors in Japan

International Nuclear Information System (INIS)

1989-01-01

The ''Long-term Program for Development and Utilization of Nuclear Energy'' by Japan Atomic Energy Commission (JAEC) plans to develop fast breeder reactors (FBRs) as the mainstay of Japan's future nuclear power generation. For the development of the FBR, the main principle is to pursue the establishment of FBR plutonium utilization systems, which are superior to the LWR uranium utilization systems. The above program also states as follows: ''Incessant efforts, with proper co-operation of the public and private sectors, will be required for a considerably long time for FBR development, and the concrete measures to be implemented in this connection will be discussed, hereafter, in continuation at the Advisory Committee on FBR Development Program of the Atomic Energy Commission''. It was clearly stated in the ''Long-term Program'' that utilities would play a leading role in the design, construction, and operation of DFBRs to follow ''Monju''. They have entrusted Japan Atomic Power Company (JAPC) with the above tasks. Technical verification and demonstration tests for seismic design and feasibility studies of main components are being conducted with government support. Regarding the nuclear fuel cycle, PNC has been developing FBR fuel reprocessing technologies. Currently, the Recycle Equipment Test Facility (RETF) is being designed to conduct engineering-scale tests. A pilot plant is planned to start operation after 2000 assuming positive results of the tests. (author)

DEMT experimental and analytical studies on seismic isolation

International Nuclear Information System (INIS)

Gantenbein, F.; Buland, P.

1989-01-01

Work on seismic isolation has been performed in France for many years, and the isolation device developed by SPIE-BATIGNOLLES in collaboration with Electricite de France (EDF) has been incorporated in the design of pressurized-water reactor (PWR) nuclear power plants. This paper reviews the experimental and theoretical studies performed at CEA/DEMT related to the overall behavior of isolated structures. The experimental work consists of the seismic shaking-table tests of a concrete cylinder isolated by neoprene sliding pads, and the vibrational tests on the reaction mass of the TAMARIS seismic facility. The analytical work consists of the development of procedures for dynamic calculation methods: for soil-structure interaction where pads are placed between an upper raft and pedestals, for time-history calculations where sliding plates are used, and for fluid-structure interaction where coupled fluid and structure motions and sloshing modes are important. Finally, this paper comments on the consequences of seismic isolation for the analysis of fast breeder reactor (FBR) vessels. The modes can no longer be considered independent (SRSS Method leads to important errors), and the sloshing increases

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Investigation on analytical method for the transfer behavior of corrosion product (CP) in the fast breeder reactors

International Nuclear Information System (INIS)

Matsuo, Youichirou; Sasaki, Shinji

2013-05-01

Radioactive corrosion products (CPs) are main cause of personal radiation exposure during maintenance work without fuel failure in FBR plants. The most important CP species are 54 Mn and 60 Co. The deposited radioactive CPs cause radiation fields near the piping and components, and the CPs contribute to the radiation exposure of the plant-worker. In this review, firstly, the collected knowledge about CP transfer behavior in the fast reactor are reviewed and analyzed. Secondly, the existing analytical methods to evaluate CP transfer behavior are investigated, issues of which and their solutions are extracted and discussed. Finally, examples of the calculated results by the improved analytical method are described. The provided conclusions are as follows; (1) Collected knowledge on CP transfer behavior. CP generation is mainly due to the dissolution of CP from hot reactor core constitution materials to hot sodium. On the core materials, particle-formed structure was confirmed. The evidence of CP precipitation in the low temperature part of the primary cooling system and the lower part of reactor core was provided. Similarly, the evidence of CP particle deposition in the same domain was also provided. (2) Extracted issues on analytical methods of CP transfer and proposed solutions. In the past, radioactivity caused by CP deposition on the piping and the core materials surface is confirmed. Subsequently, analytical models were developed based on the distribution of the CPs in the reactor coolant systems and the out-pile sodium loop test. The local high radiation dosage (such as elbow part) was observed by the radiation measurement. However, this behavior cannot be evaluated by the existing model, and it is considered necessary to take into account the transfer of CP particle. (3) The recent trend of the CP behavioral analysis method. Novel CP particle generation, transfer and deposition models were developed based on existing knowledge on CP behavior. The developed

Investigation on fuel-cladding chemical interaction in metal fuel for FBR. Reaction of rare earth elements with Fe-Cr alloy

International Nuclear Information System (INIS)

Inagaki, Kenta; Ogata, Takanari

2010-01-01

Rare-earth fission product (FP) elements generated in the metal fuel interact with cladding alloy and result in the wastage of the cladding (Fuel-Cladding Chemical Interaction (FCCI)). To evaluate FCCI quantitatively, several influential factors must be considered. They are temperature, temperature gradient, time, composition of the cladding and the behavior of rare-earth FP. In this research, the temperature and time dependencies are investigated with tests in the simplified system. Fe-12wt%Cr was used as stimulant material of cladding and rare-earth alloy 13La -24Ce -12Pr -39Nd -12Sm (RE) as a rare-earth FP. A diffusion couple Fe-Cr/RE was made and annealed at 923K, 853K, 773K or 693K. The structures of reaction layers were analyzed with Electron Probe Micro Analyzer (EPMA) and the details of the structures were clarified. The width of the reaction layer in the Fe-Cr alloy grew in proportion to the square root of time. The reaction rate constants K=(square of the width of reaction layer / time) were evaluated. It was confirmed that the relation between K and the inverse of the temperature showed linearity above 773 K. (author)

Raman distributed sensor system for temperature monitoring and leak detection in sodium circuits of FBR

Energy Technology Data Exchange (ETDEWEB)

Pandian, C.; Kasinathan, M.; Sosamma, S.; Babu Rao, C.; Jayakumar, T.; Murali, N.; Paunikar, V.; Kumar, S.; Rajan, K. K.; Raj, B. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2009-07-01

Leak detection in coolant loops of nuclear reactors is critical for the safety and performance of the reactors. The feasibility of using Raman distributed temperature sensor (RDTS) has been studied on a 30 m test loop. Temperature in sodium circuits of fast Breeder Reactor (FBR) exceeds 550 C degrees, gold coated fiber is chosen as sensor fibers. Leak is simulated through an artificial micro fissure integrated in the test loop with provision for controlled leak rate. The results are discussed in the paper. The temperature response of RDTS is compared to the conventional thermocouple and their performance was found comparable. The feasibility of detecting the temperature differential of a controlled leak with RDTS is demonstrated

Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

International Nuclear Information System (INIS)

Henderson, J.M.; Wood, S.A.; Knight, D.D.

1981-01-01

The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

Safety research experiment facilities, Idaho National Engineering Laboratory, Idaho. Final environmental impact statement

International Nuclear Information System (INIS)

Liverman, J.L.

1977-09-01

This environmental statement was prepared for the Safety Research Experiment Facilities (SAREF) Project. The purpose of the proposed project is to modify some existing facilities and provide a new test facility at the Idaho National Engineering Laboratory (INEL) for conducting fast breeder reactor (FBR) safety experiments. The SAREF Project proposal has been developed after an extensive study which identified the FBR safety research needs requiring in-reactor experiments and which evaluated the capability of various existing and new facilities to meet these needs. The proposed facilities provide for the in-reactor testing of large bundles of prototypical FBR fuel elements under a wide variety of conditions, ranging from those abnormal operating conditions which might be expected to occur during the life of an FBR power plant to the extremely low probability, hypothetical accidents used in the evaluation of some design options and in the assessment of the long-term potential risk associated with wide-acale deployment of the FBR

Analytical chemistry challenges at the back end of fuel cycle

International Nuclear Information System (INIS)

Panja, S.; Dhami, P.S.; Gandhi, P.M.

2015-01-01

Among the various nuclear fuel cycle activities, spent fuel reprocessing and nuclear waste management play key role for adaptation of closed fuel cycle option and success of three stage Indian nuclear power programme. Reprocessing mainly aims to recover fissile and fertile component from spent fuel using well known PUREX/THOREX processes. Waste management deals with all the activities which are essential for safe management of radioactive wastes that get generated during entire nuclear fuel cycle operation

Possible role of fast breeder reactor for global warming mitigation

International Nuclear Information System (INIS)

Ikeda, Kazumi; Hatano, Mamoru; Tashimo, Masanori; Sekimoto, Hiroshi

2003-01-01

Our optimization model forecasts that Fast Breeder Reactor (FBR) could be introduced as early as 2030 when nuclear power generation becomes price-competitive in the energy market. If such a scenario realizes in the future, nuclear power will be able to supply as much as 30% of total primary energy in 2100. The model also forecasts that nuclear power will contribute to reduce 6 GtC of CO 2 emission, if it completely replaces coal of the same energy. If the option of FBR is abandoned, we expect that carbon tax almost doubles in 2100 from 280 $/tC to 540 $/tC. High-burnup of LWR decreases the initial capacity of FBR plants because FBR and its fuel cycle need about seven tons of plutonium per GWe. Thus, the high-burnup of Light Water Reactor (LWR) would decrease the production of plutonium per watt-hour of electricity generation and reduce the capacity of FBR plants in the near future. However, the introduction of high breeding FBR can solve the above deficiency problem of plutonium. In other words, high breeding FBR can not only work as a catalyst for increasing the FBR plants with improved nuclear energy technology, but also contribute to reduce the greenhouse gas emission since renewable energy may be not enough in Japan. (author)

Acid in perchloroethylene scrubber solutions used in HTGR fuel preparation processes. Analytical chemistry studies

International Nuclear Information System (INIS)

Lee, D.A.

1979-02-01

Acids and corrosion products in used perchloroethylene scrubber solutions collected from HTGR fuel preparation processes have been analyzed by several analytical methods to determine the source and possible remedy of the corrosion caused by these solutions. Hydrochloric acid was found to be concentrated on the carbon particles suspended in perchloroethylene. Filtration of carbon from the scrubber solutions removed the acid corrosion source in the process equipment. Corrosion products chemisorbed on the carbon particles were identified. Filtered perchloroethylene from used scrubber solutions contained practically no acid. It is recommended that carbon particles be separated from the scrubber solutions immediately after the scrubbing process to remove the source of acid and that an inhibitor be used to prevent the hydrolysis of perchloroethylene and the formation of acids

Development of expert system for structural design of FBR components

International Nuclear Information System (INIS)

Ueda, Hiroyoshi; Uno, Masayoshi; Ogawa, Hiroshi; Shimakawa, Takashi; Yoshimura, Shinobu; Yagawa, Genki.

1995-01-01

The characteristics of structural design processes for nuclear components can be summarized as follows : (1) Many engineers belonging to different fields are working in parallel, exchanging a huge amount of data and information. (2) A final solution is determined after a number of iterative design processes. (3) Solutions have to be examined many times based on sophisticated design codes. (4) Sophisticated calculation methods such as the finite element method are frequently utilized, and experts' knowledge on such analyses plays important roles in the design process. Taking these issues into consideration, a new expert system for structural design is developed in the present study. Here, the object-oriented data flow mechanism and the blackboard model are utilized to systematize structural design processes in a computer. An automated finite element calculation module is implemented, and experts' knowledge is stored in knowledge base. In addition, a new algorithm is employed to automatically draw the design window, which is defined as an area of permissible solutions in a design parameter space. The developed system is successfully applied to obtain the design windows of four components selected from the demonstration FBR structures. (author)

Development of weld plugging for steam generator tubes of FBR

International Nuclear Information System (INIS)

Shimoyama, T.; Matsuyama, T.; Matsumoto, O.; Nagura, Y.; Nakamura, H.; Tohguchi, Y.; Kurokawa, M.; Fukada, T.

2002-01-01

This study was undertaken to develop a method of weld plugging of the heat-exchanger tubes of steam generator of Prototype FBR 'MONJU' in case these tubes are damaged for some reason. We studied mainly the shape of plug, welding procedure and effect of postweld heat treatment (PWHT). Evaporator tube sheet, tube and plug are made of 2-1/4Cr-1Mo steel and usually preheating and PWHT will be required for welding of this steel. The results of this study is as follows. 1) Plug was designed to make butt joint welding with grooved tube sheet around the tube hole to satisfy the requirements of plug designing, stress analysis, and good weldability. 2) TIG welding process was selected and certified its good weldability and good performance. 3) PWHT can be done by using high frequency induction heating method locally and also designing the plug to weld joint with tube sheet which was grooved around the tube hole. 4) Mock up test was done and it was certified that this plugging procedure has good weldability and good performance ability for Non Destructive Inspection. (author)

Improvement of numerical analysis method for FBR core characteristics. 3

International Nuclear Information System (INIS)

Takeda, Toshikazu; Yamamoto, Toshihisa; Kitada, Takanori; Katagi, Yousuke

1998-03-01

As the improvement of numerical analysis method for FBR core characteristics, studies on several topics have been conducted; multiband method, Monte Carlo perturbation and nodal transport method. This report is composed of the following three parts. Part 1: Improvement of Reaction Rate Calculation Method in the Blanket Region Based on the Multiband Method; A method was developed for precise evaluation of the reaction rate distribution in the blanket region using the multiband method. With the 3-band parameters obtained from the ordinary fitting method, major reaction rates such as U-238 capture, U-235 fission, Pu-239 fission and U-238 fission rate distributions were analyzed. Part 2: Improvement of Estimation Method for Reactivity Based on Monte-Carlo Perturbation Theory; Perturbation theory based on Monte-Carlo perturbation theory have been investigated and introduced into the calculational code. The Monte-Carlo perturbation code was applied to MONJU core and the calculational results were compared to the reference. Part 3: Improvement of Nodal Transport Calculation for Hexagonal Geometry; A method to evaluate the intra-subassembly power distribution from the nodal averaged neutron flux and surface fluxes at the node boundaries, was developed based on the transport theory. (J.P.N.)

Analytical, 1-Dimensional Impedance Model of a Composite Solid Oxide Fuel Cell Cathode

DEFF Research Database (Denmark)

Mortensen, Jakob Egeberg; Søgaard, Martin; Jacobsen, Torben

2014-01-01

An analytical, 1-dimensional impedance model for a composite solid oxide fuel cell cathode is derived. It includes geometrical parameters of the cathode, e.g., the internal surface area and the electrode thickness, and also material parameters, e.g., the surface reaction rate and the vacancy...... diffusion coefficient. The model is successfully applied to a total of 42 impedance spectra, obtained in the temperature range 555°C–852°C and in the oxygen partial pressure range 0.028 atm–1.00 atm for a cathode consisting of a 50/50 wt% mixture of (La0.6Sr0.4)0.99CoO3 − δ and Ce0.9Gd0.1O1.95 − δ...... and providing both qualitative and quantitative information on the evolution of the impedance spectra of cathodes with changing parameters....

Accelerated fuel depreciation as an economic incentive for low-leakage fuel management

International Nuclear Information System (INIS)

Downar, T.J.

1986-01-01

An analysis is presented which evaluates the tax depreciation advantage which results from the increased rate of fuel depletion achieved in the current low-leakage fuel-management LWR core reload designs. An analytical fuel-cycle cost model is used to examine the important cost parameters which are then validated using the fuel-cycle cost code CINCAS and data from the Maine Yankee PWR. Results show that low-leakage fuel management, through the tax depreciation advantage from accelerated fuel depletion, provides an improvement of several percent in fuel-cycle costs compared to traditional out-in fuel management and a constant fuel depletion rate. (author)

Critical experiment and analysis for nitride fuel fast reactor using FCA

International Nuclear Information System (INIS)

Andoh, Masaki; Iijima, Susumu; Okajima, Shigeaki; Sakurai, Takeshi; Oigawa, Hiroyuki

2000-03-01

As a research on FBR with new types of fuel, a series of experiments on a nitride fuel fast reactor was carried out at Fast Critical Assembly (FCA) to evaluate the calculation accuracy on the neutronic characteristics of the reactor. In this study, criticality, sample reactivity worth and sodium void reactivity worth were measured in the FCA XIX-2 core simulating a nitride fuel fast reactor and were analyzed using the standard analysis method for FCA fast reactor cores. The accuracy of the analysis on the effective multiplication factor was the same as those of the other FCA cores. For the plate sample reactivity worth, the calculation on the radial distribution of plutonium plate reactivity worth overestimated the measurement depending on the distance from the center of the core. For the sodium void reactivity worth, the calculation overestimated the experimental value 10 to 20% at the core center, while the overestimation was improved as the voided position was located at the core boundary. It was found that the transport effect was considerable even at the center of the core. It was considered that the calculation accuracy on the non-leakage term of the void reactivity worth and transport correction should be improved. (author)

Modification in fuel processing of Mitsubishi Nuclear Fuel's Tokai Works

International Nuclear Information System (INIS)

1976-01-01

Results of the study by the Committee for Examination of Fuel Safety, reported to the AEC of Japan, are presented, concerning safety of the modifications of Tokai Works, Mitsubishi Nuclear Fuel Co., Ltd. Safety has been confirmed thereof. The modifications covered are the following: storage facility of nuclear fuel in increase, analytical facility in transfer, fuel assemblage equipment in addition, incineration facility of combustible solid wastes in installation, experimental facility of uranium recovery in installation, and warehouse in installation. (Mori, K.)

Report on the 9th workshop on the innovative water reactor for flexible fuel cycle

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

2006-07-01

The research on Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been performed in JAEA for the development of future innovative reactor. The workshop on FLWRs has been held every year since 1998 aiming at information exchange with other organizations such as universities, laboratories, utilities and vendors. The 9th workshop was held on March 1, 2006 under the joint auspices of JAEA and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 64 participants. The workshop began with presentation entitled 'Activities on Nuclear Science and Engineering Research and Collaboration with Industry in JAEA', followed by presentations entitled 'Progress of Research and Development on FLWR' and 'On Final Report of Feasibility Study (phase 2) on Commercialized FBR Cycle Systems'. Then two lectures followed: 'Core and Fuel Design on Super Light Water Reactor' by Tokyo University and 'Recent trends on the Development of Next Generation Nuclear Reactor' by Institute of Applied Energy. This report summarizes the lectures of the workshop. (author)

Experimental and analytical analysis of polarization and water transport behaviors of hydrogen alkaline membrane fuel cell

Science.gov (United States)

Huo, Sen; Zhou, Jiaxun; Wang, Tianyou; Chen, Rui; Jiao, Kui

2018-04-01

Experimental test and analytical modeling are conducted to investigate the operating behavior of an alkaline electrolyte membrane (AEM) fuel cell fed by H2/air (or O2) and explore the effect of various operating pressures on the water transfer mechanism. According to the experimental test, the cell performance is greatly improved through increasing the operating pressure gradient from anode to cathode which leads to significant liquid water permeation through the membrane. The high frequency resistance of the A901 alkaline membrane is observed to be relatively stable as the operating pressure varies based on the electrochemical impedance spectroscopy (EIS) method. Correspondingly, based on the modeling prediction, the averaged water content in the membrane electrode assembly (MEA) does not change too much which leads to the weak variation of membrane ohmic resistance. This reveals that the performance enhancement should give the credit to better electro-chemical reaction kinetics for both the anode and cathode, also prone by the EIS results. The reversion of water back diffusion direction across the membrane is also observed through analytical solution.

Analytical chemistry department. Annual report, 1977

International Nuclear Information System (INIS)

Knox, E.M.

1978-09-01

The annual report describes the analytical methods, analyses and equipment developed or adopted for use by the Analytical Chemistry Department during 1977. The individual articles range from a several page description of development and study programs to brief one paragraph descriptions of methods adopted for use with or without some modification. This year, we have included a list of the methods incorporated into our Analytical Chemistry Methods Manual. This report is organized into laboratory sections within the Department as well as major programs within General Atomic Company. Minor programs and studies are included under Miscellaneous. The analytical and technical support activities for GAC include gamma-ray spectroscopy, radiochemistry, activation analysis, gas chromatography, atomic absorption, spectrophotometry, emission spectroscopy, x-ray diffractometry, electron microprobe, titrimetry, gravimetry, and quality control. Services are provided to all organizations throughout General Atomic Company. The major effort, however, is in support of the research and development programs within HTGR Generic Technology Programs ranging from new fuel concepts, end-of-life studies, and irradiated capsules to fuel recycle studies

3-D numerical simulation on the vibration of liquid sodium's free surface in sodium pool of FBR

International Nuclear Information System (INIS)

Han Biao; Yao Zhaohui; Ye Hongkai; Wang Xuefang

1997-01-01

This paper succeeds in simulating three-dimensional incompressible flows with free surface, complicated in-flow and out-flow boundary conditions and internal obstacles, and also can treat these fluid flows in arbitrary shape vessel using a partial cell. According to all kinds of the element influencing the free surface's vibration in sodium pool it may give the various wave's form, the highest and lowest position, and the amount of the vibration. This paper introduces the brief principle of VOF numerical method, develops the computational program based on NASA-VOF3D, provides some results about the free surface's vibration in sodium pool of FBR

The lumped parameter model for fuel pins

Energy Technology Data Exchange (ETDEWEB)

Liu, W S [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

The use of a lumped fuel-pin model in a thermal-hydraulic code is advantageous because of computational simplicity and efficiency. The model uses an averaging approach over the fuel cross section and makes some simplifying assumptions to describe the transient equations for the averaged fuel, fuel centerline and sheath temperatures. It is shown that by introducing a factor in the effective fuel conductivity, the analytical solution of the mean fuel temperature can be modified to simulate the effects of the flux depression in the heat generation rate and the variation in fuel thermal conductivity. The simplified analytical method used in the transient equation is presented. The accuracy of the lumped parameter model has been compared with the results from the finite difference method. (author). 4 refs., 2 tabs., 4 figs.

Remote maintenance system technology development for nuclear fuel cycle plants

International Nuclear Information System (INIS)

Kashihara, Hidechiyo

1984-01-01

The necessity of establishing the technology of remote maintenance, the kinds of maintenance techniques and the change, the image of a facility adopting remote maintenance canyon process, and the outline of the R and D plan to put remote maintenance canyon process in practical use are described. As the objects of development, there are twin arm type servo manipulator system, rack system, remote tube connectors, solution sampling system, inspection system for in-cell equipment, and large plugs for wall penetration. The outline of those are also reported. The development of new remote maintenance technology has been forwarded in the Tokai Works aiming at the application to a glass solidification pilot plant and a FBR fuel recycling test facility. The lowering of the rate of utilization of cells due to poor accessibility and the increase of radiation exposure of workers must be overcome to realize nuclear fuel cycle technology. The maintenance technology is classified into crane canyon method, direct maintenance cell method, remote maintenance cell method and remote maintenance canyon method, and those are described briefly. The development plan of remote maintenance technology is outlined. (Kako, I.)

Modeling the Thermal Rocket Fuel Preparation Processes in the Launch Complex Fueling System

Directory of Open Access Journals (Sweden)

A. V. Zolin

2015-01-01

Full Text Available It is necessary to carry out fuel temperature preparation for space launch vehicles using hydrocarbon propellant components. A required temperature is reached with cooling or heating hydrocarbon fuel in ground facilities fuel storages. Fuel temperature preparing processes are among the most energy-intensive and lengthy processes that require the optimal technologies and regimes of cooling (heating fuel, which can be defined using the simulation of heat exchange processes for preparing the rocket fuel.The issues of research of different technologies and simulation of cooling processes of rocket fuel with liquid nitrogen are given in [1-10]. Diagrams of temperature preparation of hydrocarbon fuel, mathematical models and characteristics of cooling fuel with its direct contact with liquid nitrogen dispersed are considered, using the numerical solution of a system of heat transfer equations, in publications [3,9].Analytical models, allowing to determine the necessary flow rate and the mass of liquid nitrogen and the cooling (heating time fuel in specific conditions and requirements, are preferred for determining design and operational characteristics of the hydrocarbon fuel cooling system.A mathematical model of the temperature preparation processes is developed. Considered characteristics of these processes are based on the analytical solutions of the equations of heat transfer and allow to define operating parameters of temperature preparation of hydrocarbon fuel in the design and operation of the filling system of launch vehicles.The paper considers a technological system to fill the launch vehicles providing the temperature preparation of hydrocarbon gases at the launch site. In this system cooling the fuel in the storage tank before filling the launch vehicle is provided by hydrocarbon fuel bubbling with liquid nitrogen. Hydrocarbon fuel is heated with a pumping station, which provides fuel circulation through the heat exchanger-heater, with

Research and development toward Monju

International Nuclear Information System (INIS)

Nakamoto, Koichiro; Ikegami, Tetsuo; Sikakura, Sakae; Iwata, Koji; Yamaguchi, Katsuhisa; Kodaira, Kiyoshi

1994-01-01

Power Reactor and Nuclear Fuel Development Corporation aimed at the development of the experimental FBR ''Joyo'' and the prototype FBR ''Monju'', and has promoted the research and development of respective fields of reactor physics, shielding, fuel, safety, sodium technology, machinery and equipment, structural materials, measurement and control by cooperating with Japan Atomic Energy Research Institute, universities, national and public research institutes, electric power and manufacturing companies. Also the promotion of the development by international cooperation has been carried out positively. As for the core and fuel, the Japan-UK joint research ''MOZART project'' for the increase of reactor power output and the heightening of fuel burnup, and the test of MOX fuel, as for machinery, equipment and structure, the rationalization of structural design, the development of steam generator, the test with sodium testing facility, and as for safety, the test on the events related to the core, decay heat removing system and containment system were carried out. (K.I.)

Review of HEDL fuel pin transient analyses analytical programs

International Nuclear Information System (INIS)

Scott, J.H.; Baars, R.E.

1975-05-01

Methods for analysis of transient fuel pin performance are described, as represented by the steady-state SIEX code and the PECT series of codes used for steady-state and transient mechanical analyses. The empirical fuel failure correlation currently in use for analysis of transient overpower accidents is described. (U.S.)

Bio Diesel An Alternative Vehicles Fuel; Analytical View

International Nuclear Information System (INIS)

El Banna, S.; El Deen, O.N.

2004-01-01

Transesterification of a vegetable oil was conducted as early as 1853, by scientists E. Duffy and J. Patrick, many years before the first diesel engine became functional(1). Rudolf Diesel's prime model, a single 10 ft (3 m) iron cylinder with a flywheel at its base, ran on its own power for the first time in Augsburg, Germany on August 10, 1893(2). Diesel later demonstrated his engine at the World Fair in Paris, France in 1898. This engine stood as an example of Diesel's vision because it was powered by peanut oil-a bio fuel. He believed that the utilization of a biomass fuel was the real future of his engine. In a 1912 speech, Rudolf Diesel said, (I) t he use of vegetable oils for engine fuels may seem insignificant today, but such oils may become, in the course of time, as important as petroleum and the coal-tar products of the present time. Rudolf Diesel was not the only inventor to believe that biomass fuels would be the mainstay of the transportation industry. Henry Ford designed his automobiles, beginning with the 1908 Model T(1), to use ethanol. Ford was so convinced that renewable resources were the key to the success of his automobiles that he built a plant to make ethanol in the Midwest and formed a partnership with Standard Oil to sell it in their distributing stations

Analytical methods used at model facility

International Nuclear Information System (INIS)

Wing, N.S.

1984-01-01

A description of analytical methods used at the model LEU Fuel Fabrication Facility is presented. The methods include gravimetric uranium analysis, isotopic analysis, fluorimetric analysis, and emission spectroscopy

Does rim microstructure formation degrade the fuel rod performance?

International Nuclear Information System (INIS)

Baron, D.; Spino, J.

2002-01-01

High burnup extension of LWR fuel is progressing to reduce the total process flow and eventually the costs of the nuclear fuel cycle. A particular fuel restructuring at high burnups, commonly observed at the periphery of LWR fuel pellets (rim structure), but also in FBR fuels to some extent and in the Plutonium rich clusters of the MOX Fuels, was considered a priori as a limitation for burnup extension. Since more than ten years this rim effect have been deeply investigated. Its causes and consequences are however not yet totally elucidated. The three steps actually identified of this phenomenon are first a progressive disappearing of the intra-granular Xenon, the outset of numerous 0.5 to 1 m pores and finally a grain subdivision around the pores. Penalty of the porosity increase on the thermal conductivity is obvious. One expect the fission gases to remain trapped in the rim porosity up to a 75 MWd/kgUO 2 local burnup. Above this threshold, 15 to 20 % of the fission gases seem to be quickly released. Microindentation tests conducted at ITU have shown the rim structure to resist fracture extension under punching. It is still open whether this implies certain ductility and viscosity of the material, or if it corresponds to stress relaxation by microcracking. Whatever the case be, it is suggested that the rim material would be able to decrease the interaction stresses and to equalise the cladding strains during a power ramp. Moreover, in the RIA tests, it was concluded so far that the grain de-cohesion caused by gas expansion at the grain boundaries was responsible for the cladding strain and failure. However, not the rim zone was affected by grain de-cohesion but the region adjacent to it. Therefore, in front of the question whether the rim structure degrades the fuel rod behaviour, we continue to argue on its benefit for fuel burnup extension. (author)

Fuels characterization studies. [jet fuels

Science.gov (United States)

Seng, G. T.; Antoine, A. C.; Flores, F. J.

1980-01-01

Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

Chernobyl reactor transient simulation study

International Nuclear Information System (INIS)

Gaber, F.A.; El Messiry, A.M.

1988-01-01

This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

Analytical Evaluation to Determine Selected PAHs in a Contaminated Soil With Type II Fuel; Metodo Optimizado de Extraccion por Ultrasonidos para la Determinacion de PAHs Seleccionados en un Suelo Contaminado con Fuel de Tipo II

Energy Technology Data Exchange (ETDEWEB)

Garcia Alonso, S.; Perez Pastor, R. M.; Sevillano Castano, M. L.; Garcia Frutos, F. J.

2010-10-21

A study on the optimization of an ultrasonic extraction method for selected PAHs determination in soil contaminated by type II fuel and by using HPLC with fluorescence detector is presented. The main objective was optimize the analytical procedure, minimizing the volume of solvent and analysis time and avoiding possible loss by evaporation. This work was carried out as part of a project that investigated a remediation process of agricultural land affected by an accidental spillage of fuel (Plan Nacional I + D + i, CTM2007-64 537). The paper is structured as: Optimization of wavelengths in the chromatographic conditions to improve resolution in the analysis of fuel samples. Optimization of the main parameters affecting in the extraction process by sonication. Comparison of results with those obtained by accelerated solvent extraction. (Author) 3 refs.

Development of system based code for integrity of FBR. Fundamental probabilistic approach, Part 1: Model calculation of creep-fatigue damage (Research report)

International Nuclear Information System (INIS)

Kawasaki, Nobuchika; Asayama, Tai

2001-09-01

Both reliability and safety have to be further improved for the successful commercialization of FBRs. At the same time, construction and operation costs need to be reduced to a same level of future LWRs. To realize compatibility among reliability, safety and, cost, the Structural Mechanics Research Group in JNC started the development of System Based Code for Integrity of FBR. This code extends the present structural design standard to include the areas of fabrication, installation, plant system design, safety design, operation and maintenance, and so on. A quantitative index is necessary to connect different partial standards in this code. Failure probability is considered as a candidate index. Therefore we decided to make a model calculation using failure probability and judge its applicability. We first investigated other probabilistic standards like ASME Code Case N-578. A probabilistic approach in the structural integrity evaluation was created based on these results, and also an evaluation flow was proposed. According to this flow, a model calculation of creep-fatigue damage was performed. This trial calculation was for a vessel in a sodium-cooled FBR. As the result of this model calculation, a crack initiation probability and a crack penetration probability were found to be effective indices. Last we discussed merits of this System Based Code, which are presented in this report. Furthermore, this report presents future development tasks. (author)

Two-dimensional analytical solution for nodal calculation of nuclear reactors

International Nuclear Information System (INIS)

Silva, Adilson C.; Pessoa, Paulo O.; Silva, Fernando C.; Martinez, Aquilino S.

2017-01-01

Highlights: • A proposal for a coarse mesh nodal method is presented. • The proposal uses the analytical solution of the two-dimensional neutrons diffusion equation. • The solution is performed homogeneous nodes with dimensions of the fuel assembly. • The solution uses four average fluxes on the node surfaces as boundary conditions. • The results show good accuracy and efficiency. - Abstract: In this paper, the two-dimensional (2D) neutron diffusion equation is analytically solved for two energy groups (2G). The spatial domain of reactor core is divided into a set of nodes with uniform nuclear parameters. To determine iteratively the multiplication factor and the neutron flux in the reactor we combine the analytical solution of the neutron diffusion equation with an iterative method known as power method. The analytical solution for different types of regions that compose the reactor is obtained, such as fuel and reflector regions. Four average fluxes in the node surfaces are used as boundary conditions for analytical solution. Discontinuity factors on the node surfaces derived from the homogenization process are applied to maintain averages reaction rates and the net current in the fuel assembly (FA). To validate the results obtained by the analytical solution a relative power density distribution in the FAs is determined from the neutron flux distribution and compared with the reference values. The results show good accuracy and efficiency.

Diesel fuel to dc power: Navy & Marine Corps Applications

Energy Technology Data Exchange (ETDEWEB)

Bloomfield, D.P. [Analytic Power Corp., Boston, MA (United States)

1996-12-31

During the past year Analytic Power has tested fuel cell stacks and diesel fuel processors for US Navy and Marine Corps applications. The units are 10 kW demonstration power plants. The USN power plant was built to demonstrate the feasibility of diesel fueled PEM fuel cell power plants for 250 kW and 2.5 MW shipboard power systems. We designed and tested a ten cell, 1 kW USMC substack and fuel processor. The complete 10 kW prototype power plant, which has application to both power and hydrogen generation, is now under construction. The USN and USMC fuel cell stacks have been tested on both actual and simulated reformate. Analytic Power has accumulated operating experience with autothermal reforming based fuel processors operating on sulfur bearing diesel fuel, jet fuel, propane and natural gas. We have also completed the design and fabrication of an advanced regenerative ATR for the USMC. One of the significant problems with small fuel processors is heat loss which limits its ability to operate with the high steam to carbon ratios required for coke free high efficiency operation. The new USMC unit specifically addresses these heat transfer issues. The advances in the mill programs have been incorporated into Analytic Power`s commercial units which are now under test.

TN-68 Spent Fuel Transport Cask Analytical Evaluation for Drop Events

International Nuclear Information System (INIS)

Shah, M.J.; Klymyshyn, Nicholas A.; Adkins, Harold E.; Koeppel, Brian J.

2007-01-01

The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing commercial spent nuclear fuel transported in casks certified by NRC under the Code of Federal Regulations (10 CFR), Title 10, Part 71 (1). Both the International Atomic Energy Agency regulations for transporting radioactive materials (2, paragraph 727), and 10 CFR 71.73 require casks to be evaluated for hypothetical accident conditions, which includes a 9-meter (m) (30-ft) drop-impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation. This paper examines the behavior of one of the NRC certified transportation casks, the TN-68 (3), for drop-impact events. The specific area examined is the behavior of the bolted connections in the cask body and the closure lid, which are significantly loaded during the hypothetical drop-impact event. Analytical work to evaluate the NRC-certified TN-68 spent fuel transport cask (3) for a 9-m (30-ft) drop-impact event on a flat, unyielding, horizontal surface, was performed using the ANSYS (4) and LS DYNA (5) finite-element analysis codes. The models were sufficiently detailed, in the areas of bolt closure interfaces and containment boundaries, to evaluate the structural integrity of the bolted connections under 9-m (30-ft) free-drop hypothetical accident conditions, as specified in 10 CFR 71.73. Evaluation of the cask for puncture, caused by a free drop through a distance of 1-m (40-in.) onto a mild steel bar mounted on a flat, essentially unyielding, horizontal surface, required by 10 CFR 71.73, was not included in the current work, and will have to be addressed in the future. Based on the analyses performed to date, it is concluded that, even though brief separation of the flange and the lid surfaces may occur under some conditions, the seals would close at the end of the drop events, because the materials remain elastic during the duration of the event

Elements of nuclear reactor fueling theory

International Nuclear Information System (INIS)

Egan, M.R.

1984-01-01

Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behavior and physical requirements of operating cycle sequences and fueling strategies having practical use in the management of nuclear fuel. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and maneuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy, and which govern fueling decisions normally made by the fuel manager. It is also demonstrated in this application that the simple batch size effect is not valid for non-integer fueling strategies, even in the simplest sequence configurations, and that it systematically underestimates the fueling requirements of degenerate sequences in general

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

International Nuclear Information System (INIS)

Reyes F, M. C.; Del Valle G, E.; Filio L, C.

2013-10-01

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Scenarios and analytical methods for UF6 releases at NRC-licensed fuel cycle facilities

International Nuclear Information System (INIS)

Siman-Tov, M.; Dykstra, J.; Holt, D.D.; Huxtable, W.P.; Just, R.A.; Williams, W.R.

1984-06-01

This report identifies and discusses potential scenarios for the accidental release of UF 6 at NRC-licensed UF 6 production and fuel fabrication facilities based on a literature review, site visits, and DOE enrichment plant experience. Analytical tools needed for evaluating source terms for such releases are discussed, and the applicability of existing methods is reviewed. Accident scenarios are discussed under the broad headings of cylinder failures, UF 6 process system failures, nuclear criticality events, and operator errors and are categorized by location, release source, phase of UF 6 prior to release, release flow characteristics, release causes, initiating events, and UF 6 inventory at risk. At least three types of releases are identified for further examination: (1) a release from a liquid-filled cylinder outdoors, (2) a release from a pigtail or cylinder in a steam chest, (3) an indoor release from either (a) a pigtail or liquid-filled cylinder or (b) other indoor source depending on facility design and operating procedures. Indoor release phenomena may be analyzed to determine input terms for a ventilation model by using a time-dependent homogeneous compartment model or a more complex hydrodynamic model if time-dependent, spatial variations in concentrations, temperature, and pressure are important. Analytical tools for modeling directed jets and explosive releases are discussed as well as some of the complex phenomena to be considered in analyzing UF 6 releases both indoors and outdoors

Amount, disposal and relative toxicity of long-lived fission products and actinides in the radioactive wastes of the nuclear fuel cycles

International Nuclear Information System (INIS)

Haug, H.O.

1975-11-01

A review is presented on the magnitude of the long-term problems of radioactive wastes from the nuclear power industry of the FRG (and Western Europe). The production of long-lived fission products and actinides has been calculated for several fuel types of the uranium-plutonium and thorium-uranium fuel cycles and related to a prediction of the development and share of LWR, FBR and HTGR. The quantities and concentrations of actinides, the radioactivity and relative toxicity index of the wastes of reprocessing (and fuel refabrication) and their changes by radioactive decay are presented. The radiotoxicity of the nuclide inventory of the solidified high-level wastes have been compared with naturally occuring uranium ores. On the long term (>10 3 years) the radiotoxicity level of the total area of the final repository in deep geological formation does not result in a significantly higher radiotoxicity level than an uranium ore deposit of low uranium content. Also discussed have been the chemical separation of the actinides from high-level wastes and recycling in fission reactors. (orig.) [de

Modeling and analytical simulation of a smouldering carbonaceous ...

African Journals Online (AJOL)

Modeling and analytical simulation of a smouldering carbonaceous rod. A.A. Mohammed, R.O. Olayiwola, M Eseyin, A.A. Wachin. Abstract. Modeling of pyrolysis and combustion in a smouldering fuel bed requires the solution of flow, heat and mass transfer through porous media. This paper presents an analytical method ...

Fundamental characterization of alternate fuel effects in continuous combustion systems

Energy Technology Data Exchange (ETDEWEB)

Blazowski, W.S.; Edelman, R.B.; Harsha, P.T.

1978-09-11

The overall objective of this contract is to assist in the development of fuel-flexible combustion systems for gas turbines as well as Rankine and Stirling cycle engines. The primary emphasis of the program is on liquid hydrocarbons produced from non-petroleum resouces. Fuel-flexible combustion systems will provide for more rapid transition of these alternate fuels into important future energy utilization centers (especially utility power generation with the combined cycle gas turbine). The specific technical objectives of the program are to develop an improved understanding of relationships between alternate fuel properties and continuous combustion system effects, and to provide analytical modeling/correlation capabilities to be used as design aids for development of fuel-tolerant combustion systems. Efforts this past year have been to evaluate experimental procedures for studying alternate fuel combustion effects and to determine current analytical capabilities for prediction of these effects. Jet Stirred Combustor studies during this period have produced new insights into soot formation in strongly backmixed systems and have provided much information for comparison with analytical predictions. The analytical effort included new applications of quasi-global modeling techniques as well as comparison of prediction with the experimental results generated.

Optimal dual-fuel propulsion for minimum inert weight or minimum fuel cost

Science.gov (United States)

Martin, J. A.

1973-01-01

An analytical investigation of single-stage vehicles with multiple propulsion phases has been conducted with the phasing optimized to minimize a general cost function. Some results are presented for linearized sizing relationships which indicate that single-stage-to-orbit, dual-fuel rocket vehicles can have lower inert weight than similar single-fuel rocket vehicles and that the advantage of dual-fuel vehicles can be increased if a dual-fuel engine is developed. The results also indicate that the optimum split can vary considerably with the choice of cost function to be minimized.

Dissolution of nuclear fuel samples for analytical purposes. I

International Nuclear Information System (INIS)

Krtil, J.

1983-01-01

Main attention is devoted to procedures for dissolving fuels based on uranium metal and its alloys, uranium oxides and carbides, plutonium metal, plutonium dioxide, plutonium carbides, mixed PuC-UC carbides and mixed oxides (PuU)O 2 . Data from the literature and experience gained with the dissolution of nuclear fuel samples at the Central Control Laboratory of the Nuclear Research Institute at Rez are given. (B.S.)

The performance of the remote analytical laboratory during the first fluorinel dissolution process campaign

International Nuclear Information System (INIS)

Lewis, L.C.; Henscheid, J.P.

1989-01-01

The Remote Analytical Laboratory at the Idaho Chemical Processing Plant was designed to provide analytical chemistry support to the irradiated fuel processing and associated waste processing operations. The facility was put into radioactive operation on July 7, 1986, and operated for more than a year during the first fluorinel fuel dissolution process campaign. The facility incorporated a number of innovative features and was equipped with state-of-the-art analytical instrumentation. The success of the facility is a direct function of how well the remote analytical equipment performed. The performance is discussed in this article

Protected plutonium breeding by transmutation of minor actinides in fast breeder reactor

International Nuclear Information System (INIS)

Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi

2008-01-01

The improvement of proliferation resistance properties of Pu and the burnup characteristics of fast breeder reactor (FBR) had been studied by utilizing minor actinides (MAs) to produce more 238 Pu from 237 Np and 241 Am through neutron capture reaction. The higher the 238 Pu content in the fuel, the higher the proliferation resistance of the fuel would be owing to the natural characteristics of 238 Pu with high decay heat and high neutron production. The present paper deals with the assessment of passive measure against nuclear material proliferation by focusing on improving the inherent proliferation barrier of discharged Pu from an FBR. Results showed that 5% MA doping to the blanket of an FBR gives as high as 17-19% 238 Pu, which could be seen as a significant improvement of the proliferation properties of Pu. Moreover, additional 5% ZrH 2 , together with 5% MA doping to the blanket, could enhance the 238 Pu fraction much more (22-24%). With an assumption of protected Pu whose 238 Pu isotopic fraction is more than 12%, the present paper revealed that protected Pu could be produced more than the Pu consumed (protected Pu breeding) through incineration in an FBR with doping of a minimum 3% MAs or (2% MAs+5% ZrH 2 ) to the blanket. (author)

A semi-analytic model of magnetized liner inertial fusion

Energy Technology Data Exchange (ETDEWEB)

McBride, Ryan D.; Slutz, Stephen A. [Sandia National Laboratories, Albuquerque, New Mexico 87185 (United States)

2015-05-15

Presented is a semi-analytic model of magnetized liner inertial fusion (MagLIF). This model accounts for several key aspects of MagLIF, including: (1) preheat of the fuel (optionally via laser absorption); (2) pulsed-power-driven liner implosion; (3) liner compressibility with an analytic equation of state, artificial viscosity, internal magnetic pressure, and ohmic heating; (4) adiabatic compression and heating of the fuel; (5) radiative losses and fuel opacity; (6) magnetic flux compression with Nernst thermoelectric losses; (7) magnetized electron and ion thermal conduction losses; (8) end losses; (9) enhanced losses due to prescribed dopant concentrations and contaminant mix; (10) deuterium-deuterium and deuterium-tritium primary fusion reactions for arbitrary deuterium to tritium fuel ratios; and (11) magnetized α-particle fuel heating. We show that this simplified model, with its transparent and accessible physics, can be used to reproduce the general 1D behavior presented throughout the original MagLIF paper [S. A. Slutz et al., Phys. Plasmas 17, 056303 (2010)]. We also discuss some important physics insights gained as a result of developing this model, such as the dependence of radiative loss rates on the radial fraction of the fuel that is preheated.

Reliability of double-wall-tube steam generator for FBR considering water leak accident frequency

International Nuclear Information System (INIS)

Ueda, Nobuyuki; Kinoshita, Izumi; Nishi, Yoshihisa

2000-01-01

For early realization, a fast breeder reactor (FBR) is required to reduce construction cost. A reactor concept in which the intermediate heat transport system is eliminated by introducing a double-wall-tube steam generator is one convincing approach. The reliability of the double-wall-tube SG in a water leak accident (sodium-water reaction accident) due to tube failure is strongly related to the mitigating system design. The safety design of the double-wall-tube SG approach is investigated to limit the accident occurrence below 10 -7 (1/ry. A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to effectively prevent adhesion of the double-wall-tube. The reliability of the tube-to-tube plate was evaluated at 10 -10 (l/hr) for an inner tube and 10 -9 (l/hr) for an outer with reference to the failure experience of previous SGs. The failure must be detected within 30 to 60 minutes. (author)

Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

2016-05-15

U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

Safety Research Experiment Facilities, Idaho National Engineering Laboratory, Idaho. Draft environmental statement

International Nuclear Information System (INIS)

1977-01-01

This environmental statement was prepared in accordance with the National Environmental Policy Act of 1969 (NEPA) in support of the Energy Research and Development Administration's (ERDA) proposal for legislative authorization and appropriations for the Safety Research Experiment Facilities (SAREF) Project. The purpose of the proposed project is to modify some existing facilities and provide a new test facility at the Idaho National Engineering Laboratory (INEL) for conducting fast breeder reactor (FBR) safety experiments. The SAREF Project proposal has been developed after an extensive study which identified the FBR safety research needs requiring in-reactor experiments and which evaluated the capability of various existing and new facilities to meet these needs. The proposed facilities provide for the in-reactor testing of large bundles of prototypical FBR fuel elements under a wide variety of conditions, ranging from those abnormal operating conditions which might be expected to occur during the life of an FBR power plant to the extremely low probability, hypothetical accidents used in the evalution of some design options and in the assessment of the long-term potential risk associated with wide-scale deployment of the FBR

Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

International Nuclear Information System (INIS)

Goswami, S.

1983-01-01

The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

Enforcement of evaluation by achievement analysis system

International Nuclear Information System (INIS)

Konishi, Yasutoshi; Sonoyama, Minoru; Suzuki, Atsushi

2004-02-01

Japan Nuclear Cycle Development Institute (JNC) has developed FBR achievement analysis system by the last fiscal year and has enforced the investigation of its functional expansion. That system is based on the AHP (Analytic Hierarchy Process) to do comparative evaluation multilaterally between proposed concepts of FBR cycle or between FBR cycle and other power source systems. This fiscal year, we enforced achievement analysis for 22 cases of proposed concepts of FBR cycle and between FBR cycle and other power source systems (LWR, thermal power generation, hydraulic power generation, etc.). The evaluation items related with technical feasibility and social acceptability were included in addition to those of economy, resource utilization effectiveness, environmental burden reduction, nuclear proliferation resistance and safety. Also, we investigated social changes that could happen in our country in the future, and we drew 4 future scenarios combining likely changes, then we investigated classifications of weight that seem to be adequate under each scenario with its calculation logic. In establishing points of view or structure of evaluation, and in the process of drawing scenarios, we collected comments from experts in OR (Operations Research) field and energy field. (author)

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania's 14-MW TRIGA reactor

International Nuclear Information System (INIS)

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-01-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations

Earthquake-proof support structures for the recycling pump in FBR type reactors

International Nuclear Information System (INIS)

Nakagawa, Masaki; Shigeta, Masayuki.

1984-01-01

Purpose: To improve the earthquake proofness of the recycling pump for use in FBR type reactors upon earthquake by reducing the vibration response of the pump. Constitution: The outer casing of a recycle pump suspended into liquid sodium is extended to the portion that penetrates a reactor core support structures. Support structures surrounding the outer side of the recycling pump are disposed with a gap not restraining the free thermal deformations of the recycling pump to the inside of the partition wall structures and the portion of the recycling pump penetrating the reator core support structures, to integrate the support structures with the reactor core support structures. Accordingly, there are no interferences between the recycling pump and the support structures with respect to the thermal deformations that change gradually with time. Upon vibrating under the rapidly changing external forces of earthquakes, however, the pressure resulted to the liquid in the gap due to the vibrations of the recycling pump is transmitted with no escape to the support structures, the recycling pump and the support structures integrally resist the vibrations thereby enabling to reduce the vibrations in the recycling pumps. (Horiuchi, T.)

Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

International Nuclear Information System (INIS)

Sato, Ikken; Lemoine, Francette; Struwe, Dankward

2004-01-01

In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear density and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long

Elements of nuclear reactor fueling theory

International Nuclear Information System (INIS)

Egan, M.R.

1984-01-01

Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behaviour and physical requirements of operating cycle sequences and fueling strategies having practical use in fuel management. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and manoeuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy. Numerical evaluations of degenerate equilibrium cycle sequences are then performed for a typical PWR core, and accompanying fuel cycle costs are calculated. The impact of design and operational limits as constraints on the performance mappings for this reactor are also studied with respect to achieving improved cost performance from the once-through fuel cycle. The dynamics of transition cycle sequences are then examined using the generalized theory. Proof of the existence of non-degenerate equilibrium cycle sequences is presented when the mechanics of the fixed reload batch size strategy are developed analytically for transition sequences. Finally, an analysis of the fixed reload enrichment strategy demonstrates the potential for convergence of the transition sequence to a fully degenerate equilibrium sequence. (author)

Analysis of transient fuel failure mechanisms: selected ANL programs

International Nuclear Information System (INIS)

Deitrich, L.W.

1975-01-01

Analytical programs at Argonne National Laboratory related to fuel pin failure mechanisms in fast-reactor accident transients are described. The studies include transient fuel pin mechanics, mechanics of unclad fuel, and mechanical effects concerning potential fuel failure propagation. (U.S.).

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)