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Sample records for mk-ii joyo mk-ii

  1. JOYO MK-II core characteristics database

    Tabuchi, Shiro; Aoyama, Takafumi; Nagasaki, Hideaki; Kato, Yuichi

    1998-12-01

    The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, extensive data were accumulated from the core characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database. The code system MAGI has been developed and used for core management of JOYO MK-II, and the core characteristics and the irradiation test conditions were calculated using MAGI on the basis of three dimensional diffusion theory with seven neutron energy groups. The core management data include extensive data, which were recorded on CD-ROM for user convenience. The data are specifications and configurations of the core, and for about 300 driver fuel subassemblies and about 60 uninstrumented irradiation subassemblies are core composition before and after irradiation, neutron flux, neutron fluences, fuel and control rod burn-up, and temperature and power distributions. MK-II core characteristics and test conditions were stored in the database for post analysis. Core characteristics data include excess reactivities, control rod worths, and reactivity coefficients, e.g., temperature, power and burn-up. Test conditions include both measured and calculated data for irradiation conditions. (author)

  2. JOYO MK-II core characteristics database

    Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Aoyama, Takafumi

    2003-04-01

    The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worth, reactor kinetic parameters and the MK-II core performance test results were included per user's requests. The core characteristics obtained from the 32 nd to 35 th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded to CD-ROM for user convenience. The Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. (author)

  3. Development of JOYO MK-II core characteristics database

    Tabuchi, Shiro; Aoyama, Takafumi

    2000-01-01

    The MK-II core of the experimental fast reactor JOYO served as the irradiation bed for testing fuels and materials for FBR development since 1982 for 15 years. During the MK-II operation, extensive data were accumulated from the core management calculations and characteristics tests conducted in thirty-one duty operations and thirteen special test operations. These core management data and core characteristics data were compiled into a database recorded on CD-ROM for user convenience. The calculated core management data are the text style data. The 'Configuration Data' include the history of the fuel exchange and core arrangement for each cycle. The Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of about 300 fuel subassemblies, and 60 irradiation subassemblies. The 'Output Data' include the neutron fluxes, gamma fluxes, power density, linear heat rates, coolant and fuel temperature distributions of each core position at the beginning and end of each cycle. The measured core characteristics data, such as the excess reactivity, control rod worths, temperature coefficient, power coefficient, and burn-up coefficient are also included along with the measurement conditions. (J.P.N.)

  4. JOYO coolant sodium and cover gas purity control database (MK-II core)

    Ito, Kazuhiro; Nemoto, Masaaki

    2000-03-01

    The experimental fast reactor 'JOYO' served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon dioxide, methane and helium in argon gas with the reactor condition. (author)

  5. JOYO MK-II core characteristics database. Update to JFS-3-J3.2R

    Ohkawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi

    2003-04-01

    The 'JOYO' MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition in 2001. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. However, after the database was published, it was recently found that there were errors in the process of making the group constant set JFS-3-J3.2, and it was revised at JFS-3-J3.2R. Then, the group constant set was updated at JFS-3-J3.2R in this database. The MK-II core management data nad core characteristics data were recorded on CD-ROM for user convenience. The structure of the database is the same as in the first edition. The 'Configuration Data' include the core arrangement and refueling record for each operational cycle. The 'Subassembly Library Data' include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The 'Output Data' contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The 'Core Characteristics Data' include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. The effect of updating the group constant set, the calculation results of excess reactivity decreased by about 0.15Δk/kk', and the effects to other core characteristics were negligible. (author)

  6. Postirradiation examination of JOYO MK-II control rod (CRM601). Irradiation performance of shroud type absorber pin

    Tanaka, Kosuke; Kikuchi, Shin; Katsuyama, Kozo; Nagamine, Tsuyoshi; Mitsugi, Takeshi; Uto, Manabu; Tatebe, Kazuaki; Onose, Shoji; Maruyama, Tadashi

    1998-10-01

    This paper describes the results of postirradiation examination and analysis by CORAL code for irradiation performance of CRM601 control rod, which was the 6th reloaded control rod with shroud type absorber pins for use in JOYO MK-II core. The detailed visual examination indicated that there was no cladding breach in absorber pins. However, sodium ingress from the vent tube was observed in four absorber pins among seven pins. While a remarkable oval deformation occurred in cladding tube of helium bonded absorber pins, a little or no diametral change was observed in the absorber pins in which sodium ingress took place. From metallurgical observations and the analysis by CORAL code, it was estimated that the shroud tube installed in helium bonded absorber pins were irradiated at 720degC, and those in sodium bonded absorber pins were irradiated at 420degC. It was confirmed that diametral change of cladding depended on the initial gap between shroud and cladding tube. The results of present investigation indicate that it is desirable to use the materials with low thermal expansion coefficient for shroud tubes, and that sodium bonded absorber pins were advantageous for obtaining long life control rods. (author)

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

    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)

  8. Resurfacing the Jodrell Bank Mk II radio telescope

    Spencer, R. E.; Haggis, J. S.; Morrison, I.; Davis, R. J.; Melling, R. J.

    The improvement of the short-wavelength performance of the Jodrell Bank Mk II radio telescope is described. A final rms profile error of 0.6 mm was achieved due to the invention of an inexpensive technique of panel construction and measurement combined with the use of radio-astronomical holographic techniques to measure the telescope under actual operating conditions. Some further improvements to extend the short wavelength performance are suggested.

  9. Jet Propellant (JP)-8 Fuel Evaluation Test Mk II - Reset (Mk II R) Bridge Erection Boat (BEB)

    2008-10-01

    diesel engines (fig. 2 and 3) equipped with Delphi rotary fuel injection pumps. Figure 1. Mk II R BEB pushing a two-bay IRB raft. TR No. WF-E-83 2... nozzles . The new pump (serial No. 08813K7B) and gasket were installed. 24 May 07 51.0 50.4 44.9 103 Port Fuel Pump and Injectors Replaced. At the...part No. 3909356) were installed on the injector nozzles . The new pump (serial No. 59640HZB) and gasket were installed. 31 May 07 51.5 50.5 44.9 104

  10. Reference design (MK-I and MK-II) for experimental multi-purpose VHTR

    Miyamoto, Yoshiaki; Suzuki, Kunihiko; Sato, Sadao

    1975-10-01

    This report summarizes the results of a study on thermal and mechanical performances of the core, which are obtained in course of reference design (Mk-I and Mk-II) for the experimental multi-purpose VHTR: (1) Design criteria, design methods and design data. These bases are also discussed in order to refer in the case of proceeding a next design work. (2) The results of performance analysis such as the initial core and its prediction for the irradiated core. (auth.)

  11. Operating experience of TRIGA MK-II Research Reactor in Bangladesh

    Mannan, M.A.; Ahmed, K.

    1992-01-01

    A 3 MW TRIGA MK II Research Reactor was installed in Bangladesh in 1986. The reactor is being utilized for research, training and for production of radioisotopes. Recently two faults were detected, one in the Emergency Core Cooling System and the other in the Primary Coolant Loop, which hindered the operation of the reactor partially. The faults were investigated by a team of local experts. Results of analyses of possible initiating events of the faults and the remedial steps are briefly discussed in the paper. (author)

  12. Application of bone scintigrams in total knee replacement (Okayama MK-II type)

    Umeda, T.; Inoue, S.; Matsui, N.; Moriya, H. (Chiba Univ. (Japan). School of Medicine)

    1982-02-01

    Eighteen patients with 21 total knee replacements (OKAYAMA MK-II type) were examined by radionuclide imaging in order to assess the prosthetic complaints such as loosening, infection, fracture and lasting pain. The following results and conclusions were obtained. 1) Bone imaging can reveal the condition of the attachment of bone and prosthesis. 2) Diffuse uptake gradually diminished until 18 months after surgery. 3) In front view on bone imaging, tibial uptake corresponded highly with the part of the weight area. 4) In cases of high uptake of posterior femoral component in lateral view, the range of knee flexion was mostly restricted. 5) Long-period persistent local uptake suggested loosening of the prosthesis or fracture of the tibial plateau. 6) Patello-femoral uptake showed no relation to the patellofemoral complaints. Radionuclide bone imaging is considered to represent one of the most valuable diagnostic procedures for assessing the clinical results after total knee replacement.

  13. Application of bone scintigrams in total knee replacement (Okayama MK-II type)

    Umeda, Tohru; Inoue, Shunichi; Matsui, Nobuo; Moriya, Hideshige

    1982-01-01

    Eighteen patients with 21 total knee replacements (OKAYAMA MK-II type) were examined by radionuclide imaging in order to assess the prosthetic complaints such as loosening, infection, fracture and lasting pain. The following results and conclusions were obtained. 1) Bone imaging can reveal the condition of the attachment of bone and prosthesis. 2) Diffuse uptake gradually diminished until 18 months after surgery. 3) In front view on bone imaging, tibial uptake corresponded highly with the part of the weight area. 4) In cases of high uptake of posterior femoral component in lateral view, the range of knee flexion was mostly restricted. 5) Long-period persistent local uptake suggested loosening of the prosthesis or fracture of the tibial plateau. 6) Patello-femoral uptake showed no relation to the patellofemoral complaints. Radionuclide bone imaging is considered to represent one of the most valuable diagnostic procedures for assessing the clinical results after total knee replacement. (author)

  14. MOTHER MK II: An advanced direct cycle high temperature gas reactor

    Hart, R.S.; Kendall, J.M.; Marsden, B.J.

    2003-01-01

    -power refueling feature of the Pebble Bed reactor core concept is attractive in many situations, the MOTHER MK II conceptual design adopts a Pebble Bed core configuration. The power conversion systems of MOTHER MKI are utilized. In an effort to overcome the disadvantages of current graphite pebble annular Pebble Bed core designs, MOTHER MK II introduces a novel split core configuration. The MOTHER concepts were developed with an objective of minimizing technical risk and the need for technology development. A principal purpose of this paper is to inform other designers currently working on direct cycle HTGR concepts of the work undertaken in defining the designs for the MOTHER nuclear power plants, and of the many novel technical features adopted. (author)

  15. Assessment of gold flux monitor at irradiation facilities of MINT TRIGA MK II reactor

    Wee Boon Siong; Abdul Khalik Wood; Mohd Suhaimi Hamzah; Shamsiah Abdul Rahman; Md Suhaimi Elias; Nazaratul Ashifa Abd Salim

    2005-01-01

    Neutron source of MINTs TRIGA MK II reactor has been used for activation analysis for many years and neutron flux plays important role in activation of samples at various positions. Currently, two irradiation facilities namely the pneumatic transfer system and rotary rack are available to cater for short and long lived irradiation. Neutron flux variation for both irradiation facilities have been determined using gold wire and gold solution as flux monitor. However, the use of gold wire as flux monitor is costlier if compared to gold solution. The results from analysis of certified reference materials showed that gold solution as flux monitors yield satisfactory results and proved to safe cost on the purchasing of gold wire. Further experiment on self-shielding effects of gold solution at various concentrations has been carried out. This study is crucial in providing vital information on the suitable concentration for gold solution as flux monitor. In the near future, gold solution flux monitor will be applied for routine analysis and hence to improve the capability of the laboratory on neutron activation analysis. (Author)

  16. Assessing the Frequency and Material Consequences of Collisions with Vessels Lying at an Anchorage in Line with IALA iWrap MkII

    Hans-Christoph Burmeister

    2014-03-01

    Full Text Available This paper proposes a collision model for ships underway and temporary objects as an extension to state-of-the-art maritime risk assessment like IALA iWrap MkII. It gives a brief review of frequency modeling's and consequence calculation theory as well as its applications, before it analogously derives a model to assess the risk of anchorage areas. Subsequently, its benefit is demonstrated by an example scenario.

  17. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  18. In-core flow rate distribution measurement test of the JOYO irradiation core

    Suzuki, Toshihiro; Isozaki, Kazunori; Suzuki, Soju

    1996-01-01

    A flow rate distribution measurement test was carried out for the JOYO irradiation core (the MK-II core) after the 29th duty cycle operation. The main object of the test is to confirm the proper flow rate distribution at the final phase of the MK-II core. The each flow rate at the outlet of subassemblies was measured by the permanent magnetic flowmeter inserted avail of fuel exchange hole in the rotating plug. This is third test in the MK-II core, after 10 years absence from the final test (1985). Total of 550 subassemblies were exchanged and accumulated reactor operation time reached up to 38,000 hours from the previous test. As a conclusion, it confirmed that the flow rate distribution has been kept suitable in the final phase of the MK-II core. (author)

  19. Upgrading program of the experimental fast reactor Joyo

    Yoshida, A.; Yogo, S.

    2001-01-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  20. Bowing behavior of subassemblies in experimental fast reactor ''JOYO''

    Ikegami, T.; Mizoo, N.; Matsuno, Y.; Watari, Y.

    1984-01-01

    In JOYO, the measured power coefficients in the beginning of the operation cycle of MK-I and MK-II cores showed power dependence, while the calculation without taking account of bowing predicted little power dependence. The bowing analysis was performed in order to investigate the power dependence observed in the measured power coefficients and the following conclusions were obtained. (1) The evaluated power coefficients taking account of bowing effect agree better with measured ones than the calculated ones without taking account of bowing effect in MK-I core. (2) In MK-II core, although the analytical results show not so good agreement quantitatively with the measured power coefficients, it is suggested that they agree better depending on the uncertain parameters such as the heat generation in the reflector region, the threshold moment for leaning and the stiffness of the inner reflector. (3) It becomes clear from these results that the power dependence observed in the measured power coefficients in JOYO is due to the bowing effect. (author)

  1. Development of the flow control irradiation facility for JOYO

    Soroi, Masatoshi; Miyakawa, Shun-ichi

    1998-05-01

    This report describes the present situation and problems with the development of the flow control irradiation facility (FLORA). The purpose of FLORA is to run the cladding breach (RTCB) irradiation test under loss of flow conditions in the experimental fast reactor 'JOYO'. FLORA is a facility like FPTF (Fuel Performance Test Facility) plus BFTF (Breached Fuel Test Facility) in EBR-II, USA. The technical feature of FLORA is its annular linear induction pump (A-LIP), which was developed in response to a need identified through the experiences in the mechanical flow control of FPTF. We have already designed the basic system facility of FLORA for the JOYO MK-II core. However, to put FLORA to practical use in the future, we have to confirm the stability of the JOYO MK-III core condition, solve problems and improve the design. We are going to freeze and review the FLORA project, taking into consideration the fuel development situation and the research project of JOYO MK-III core. (J.P.N.)

  2. Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

    Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

    2012-10-01

    A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

  3. Analysis of excess reactivity of JOYO MK-III performance test core

    Maeda, Shigetaka; Yokoyama, Kenji

    2003-10-01

    JOYO is currently being upgraded to the high performance irradiation bed JOYO MK-III core'. The MK-III core is divided into two fuel regions with different plutonium contents. To obtain a higher neutron flux, the active core height was reduced from 55 cm to 50 cm. The reflector subassemblies were replaced by shielding subassemblies in the outer two rows. Twenty of the MK-III outer core fuel subassemblies in the performance test core were partially burned in the transition core. Four irradiation test rigs, which do not contain any fuel material, were loaded in the center of the performance test core. In order to evaluate the excess reactivity of MK-III performance test core accurately, we evaluated it by applying not only the JOYO MK-II core management code system MAGI, but also the MK-III core management code system HESTIA, the JUPITER standard analysis method and the Monte Carlo method with JFS-3-J3.2R content set. The excess reactivity evaluations obtained by the JUPITER standard analysis method were corrected to results based on transport theory with zero mesh-size in space and angle. A bias factor based on the MK-II 35th core, which sensitivity was similar to MK-III performance test core's, was also applied, except in the case where an adjusted nuclear cross-section library was used. Exact three-dimensional, pin-by-pin geometry and continuous-energy cross sections were used in the Monte Carlo calculation. The estimated error components associated with cross-sections, methods correction factors and the bias factor were combined based on Takeda's theory. Those independently calculated values agree well and range from 2.8 to 3.4%Δk/kk'. The calculation result of the MK-III core management code system HESTLA was 3.13% Δk/kk'. The estimated errors for bias method range from 0.1 to 0.2%Δk/kk'. The error in the case using adjusted cross-section was 0.3%Δk/kk'. (author)

  4. Japan: The Experimental Fast Reactor JOYO. Profile 12

    2017-01-01

    The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

  5. Integral test of JENDL dosimetry file using fast neutron field in the Experimental Fast Reactor JOYO

    Aoyama, Takafumi; Sekine, Takashi

    1999-09-01

    In order to evaluate the applicability of the JENDL dosimetry file, an integral test using a fast neutron spectrum field in the Experimental Fast Reactor JOYO Mark-II core was performed. The dosimeter set consisting of eight reactions of 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 58 Fe(n,γ) 59 Fe, 58 Ni(n,p) 58 Co, 59 Co(n,γ) 60 Co, 63 Cu(n,α) 60 Co, 238 U fission and 237 Np fission was irradiated for approximately 30 days near the core center of the JOYO Mk-II. Neutron flux at the dosimeter position was calculated using the two dimensional discrete ordinate transport code 'DORT'. The core configuration was modeled in XY geometry, and the 100 group cross section set of JSD-J2 / JFT-J2, which was processed from JENDL-2, was utilized. The absolute value of neutron flux was normalized so that the 235 U fission rate using the calculated neutron spectrum agreed with the measured reaction rate. The 103 group cross section data were processed by 'NJOY' code for nuclides to be used in the JOYO dosimetry. As the results of integral test for JENDL/D-99 (new file) and JENDL/D-91 (previous file), calculated values by JENDL/D-99 agreed well with the experimental values, and the C/E ratios ranged from 0.95 to 1.22. By comparing the results between JENDL/D-99 and JENDL/D-91, small differences exist, except for 58 Fe(n, γ) 59 Fe reaction, which was improved significantly in JENDL/D-99. (author)

  6. Deuterium to helium plasma-wall change-over experiments in the JET MkII-gas box divertor

    Hillis, D.L.; Loarer, T.; Bucalossi, J.; Pospieszczyk, A.; Fundamenski, W.; Matthews, G.; Meigs, A.; Morgan, P.; Phillips, V.; Pitts, R.; Stamp, M.; Hellermann, M. von

    2003-01-01

    The deuterium and helium dynamics in the plasma and subdivertor regions of JET are compared during a sequence of similar ohmic and ICRH pulses where 100% He gas is injected into the JET vacuum vessel, whose graphite walls were previously saturated with deuterium. After the first six He fueled change-over discharges, only He plasma operation was performed. Following this investigation, the situation is reversed and the change-over from an initially saturated He wall is investigated when only D 2 plasma fuelling is used. The He concentration is measured in the subdivertor with a species selective Penning gauge. Comparison of the time dependence of the divertor concentrations with those at the edge and strike point shows significant differences during the first six discharges. This difference along with a global He particle balance is used to assess the status of the wall saturation over the initial 6-7 He change-over discharges

  7. Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

    Guthrie, J.A.S.; Harding, N.H.

    1981-01-01

    This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

  8. Design and manufacture of mechanical forceps to pick up objects at the bottom of the pool reactor TRIGA MK II

    Kankunku, K.P.; Lukanda, M.V.

    2011-01-01

    This design helps us to pick up any objects felt in bottom of swimming pool, which is a radioactive area, due to the presence of spent nuclear fuel. Its great advantage is its sample designing and made with local material.

  9. JOYO MK-III performance test. Criticality test, excess reactivity measurement and burn-up coefficient measurement

    Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki

    2005-03-01

    The MK-III performance test began in June 2003 to fully characterize the upgraded core and heat transfer system of the experimental fast reactor JOYO. This paper describes the results of the approach to criticality, the excess reactivity evaluation and the burn-up coefficient measurement. In the approach to criticality test, the MK-III core achieved initial criticality at the control rod bank position of 412.8 mm on 14:03 July 2nd, 2003. Because the replacement of the outer two rows of reflector subassemblies with shielding subassemblies reduced the source range monitor signals by a factor of 3 at the same reactor power compared with those in the MK-II core, we measured the change of the monitor's response and determined the count rate 2x10 4 cps.' as an appropriate value judging the zero power criticality. In the excess reactivity evaluation, the zero power excess reactivity at 250degC was 2.99±0.10%Δk/kk' based on the measured critical rod bank position and the measured control rod worths. The predicted value by the JOYO core management code system HESTIA was 3.13±0.16%Δk/kk', showing good agreement with the measured value. The measured excess reactivity was within the safety requirement limit. In the burn-up coefficient measurement, the excess reactivity change versus the reactor burn-up was evaluated. The measurement method adopted was to measure the control rod positions during the rated power operation. A value of -2.12x10 -4 Δk/kk'/MWd was obtained as a measured burn-up coefficient. The value calculated by HESTIA was -2.12x10 -4 Δk/kk'/MWd, and it agreed well with the measured value. All technical safety requirements for MK-III core were satisfied and the calculation accuracy of the core management code system HESTIA was confirmed. (author)

  10. MK-III function tests in JOYO. Dump heat exchanger (DHX)

    Kawahara, Hirotaka; Isozaki, Kazunori; Ishii, Takayuki; Ichige, Satoshi; Sakaba, Hideo; Nakai, Satoru; Nose, Shouiti

    2004-06-01

    A key part of the upgrade of the experimental fast reactor JOYO to the MK-III design was the replacement of the dump heat exchangers. MK-III function tests (SKS-1) of the new dump heat exchangers were carried out from August 27, 2001 through September 13, 2001. The major results of the function tests of the dump heat exchangers were as follows: (1) Air flow of the main blower with an inlet vane opening of 50% was confirmed to exceed the design rated flow of 7,700 m 3 /min. It was also demonstrated that an inlet vane opening of 100% provides about 130% of the design rated flow. This is because the new DHX flow route has more low pressure loss than the design value. (2) Tests of the air flow of the main blower demonstrated that with a fully opened inlet damper, a full opened outlet damper and an inlet vane opening of 0% provides about 5% of the design rated flow. (3) Free flow coast down characteristics of the main blower achieved an inlet vane 0% opening in an average of 7.9 seconds. Revolutions per minute of the main blower reached zero in an average of 8.7 seconds. The delay time from the opening of the vacuum contact breaker to the air flow decrease was approximately 1 second. This was a more conservative value than the 5 seconds assumed in design thermal transient analyses. (4) The loudest noise occurred with the main blower operating with a 25% inlet vane opening. At that time, the noise around the main blower was approximately 100 dB, and in the surrounding monitoring area boundary, the noise was 50 dB. This was confirmed to be within the standard of the Ibaraki prefectural ordinance. (5) Although the MK-III inlet vane and inlet damper drive unit was bigger than the MK-II unit, the accumulator tank was confirmed to provide sufficient volume during a compression air loss event. (author)

  11. Joyo progress report, vol. 8

    1983-01-01

    Following Joyo Reactor Technology Progress Reports (Vol. 1 to Vol. 7), the name was changed to Joyo Progress Report from this volume, and the activities concerning the fast breeder experimental reactor Joyo as a whole are to be reported as quarterly report. In the fast breeder experimental reactor Joyo, the change to the core for irradiation (MK-2) from the core for breeding (MK-1) was carried out since January, 1982, in order to utilize the reactor as an irradiation facility for the development of fuel and materials. The main work was the construction of the core for irradiation by exchanging 290 fuel elements, and the exchange of upper and lower guide pipes for control rods, the reconstruction of the driving mechanism, the installation of standby neutron detector system, the acceptance and inspection of new fuel, and the transfer of spent fuel between pools were carried out. As scheduled, the core for irradiation attained the initial criticality on November 22, and the works of constructing the core were completed on December 23, 1982. Thereafter, the 100 MW performance test was begun. Various experience and valuable data were obtained in the regular inspection and the maintenance and repair works carried out at the same time, regarding the operation and maintenance of the Joyo facilities. (Kako, I.)

  12. WATER DEPTH - AVERAGE and Reflectance- Intensity collected from LADS Mk II Airborne System in Caribbean Sea and Puerto Rico from 2006-04-07 to 2006-05-15 (NCEI Accession 0153360)

    National Oceanic and Atmospheric Administration, Department of Commerce — These images represent LiDAR (Light Detection & Ranging) data collected by NOAA from the shoreline of southwestern Puerto Rico to about 50 meters in depth....

  13. The JOYO remote monitoring system

    Damico, Joseph P.; Hashimoto, Yu

    2000-01-01

    The evolution of the personal computer, operating systems and applications software and the Internet has brought drastic change and many benefits worldwide. Remote monitoring systems benefit from computer network and other modern software technologies. The availability of fast, inexpensive and secure communications enables new solutions for monitoring system applications. The JOYO Remote Monitoring System (RMS) utilizes computer network communications and modular software design to provide a distributed integrated solution for monitoring multiple storage locations. This paper describes the remote monitoring system installed at the JOYO Fast Reactor. The system combines sensors, software, and computer network technologies to create a powerful data collection, storage and dissemination capability. The RMS provides a flexible, scalable solution for a variety of applications. The RMS integrates a variety of state of the art technologies from several sources and serves as a test bed for cutting edge technologies that can be shared with outside users. This paper describes the system components and their operation and discusses system benefits. Current activities and future plants for the JOYO RMS will be discussed. (author)

  14. Introduction of the experimental fast reactor JOYO

    Matsuba, Ken-ichi; Kawahara, Hirotaka; Aoyama, Takafumi

    2006-01-01

    The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO. (author)

  15. Fuel pellet relocation behavior in fast reactor uranium-plutonium mixed oxide fuel pin at beginning-of-life

    Inoue, Masaki; Ukai, Shigeharu; Asaga, Takeo

    1999-08-01

    The effects of fabrication parameters, irradiation conditions and fuel microstructural feature on fuel pellet relocation behavior in fast reactor fuel pins were investigated. This work focused only on beginning-of-life conditions, when fuel centerline temperature depends largely on the behavior. Fuel pellet relocation behavior in Joyo Mk-II driver could not be characterized because of the lack of data. And the behavior in FFTF driver and its larger diameter type fuel pins could not be characterized because of the extensive lot-by-lot scatters. The behavior both in Monju type and in Joyo power-to-melt type fuel pins were similar to each other, and depends largely on the as-fabricated gap width while the effects of linear heat rate and the extent of microstructural evolution were negligible. And fuel pellet centerline melting seems to affect slightly the behavior. The correlation, which describes the extent of relocation both in Monju type and in Joyo power-to-melt type fuel pins, were newly formulated and extrapolated for Joyo Mk-II driver, FFTF driver and its larger diameter type fuel pins. And the behavior in Joyo Mk-II driver seemed to be similar. On the contrary, the similarity with JNC fuel pins was observed case-by-case in FFTF driver and its larger diameter type fuel pins. (author)

  16. JOYO construction and preoperational test experience

    1976-03-01

    The construction and installation of Joyo, the first experimental fast reactor in Japan, have been completed. The application for the license for Joyo was made with the output of 50 MWt, and the power raising to 100 MWt target is left for future. Joyo is a sodium-cooled fast breeder reactor with mixed oxide fuel. The research and development for and the construction of Joyo are described. The initial stage of the preoperational test has been finished, and the further test stage is in progress. Sodium-cooled fast reactors are operated at higher temperature and lower pressure as compared with light water reactors, therefore the thermal stress is high, while the mechanical stress is low. The materials used for the sodium components are apt to creep, while the mechanical properties are impaired by the mass transfer in the hot sodium circuits. The guide lines for the structural design of Joyo were established on the basis of ASME Boiler and Pressure Vessel Code. The basic philosophy and the method of the aseismatic design for Joyo are almost same as those for large commercial reactor plants. The thermal shock due to air blast coolers cannot be avoided in LMFBRs, but care should be taken in the design to mitigate the shock. It is desirable to establish more detailed standards on inspection and examination to cope with complex LMFBRs. (Kako, I.)

  17. Power coefficient anomaly in JOYO

    Yamamoto, H

    1980-12-15

    Operation of the JOYO experimental fast reactor with the MK-I core has been divided into two phases: (1) 50 MWt power ascension and operation; and (2) 75 MWt power ascension and operation. The 50 MWt power-up tests were conducted in August 1978. In these tests, the measured reactivity loss due to power increases from 15 MWt to 50 MWt was 0.28% ..delta.. K/K, and agreed well with the predicted value of 0.27% ..delta.. K/K. The 75 MWt power ascension tests were conducted in July-August 1979. In the process of the first power increase above 50 MWt to 65 MWt conducted on July 11, 1979, an anomalously large negative power coefficient was observed. The value was about twice the power coefficient values measured in the tests below 50 MW. In order to reproduce the anomaly, the reactor power was decreased and again increased up to the maximum power of 65 MWt. However, the large negative power coefficient was not observed at this time. In the succeeding power increase from 65 MWt to 75 MWt, a similar anomalous power coefficient was again observed. This anomaly disappeared in the subsequent power ascensions to 75 MWt, and the magnitude of the power coefficient gradually decreased with power cycles above the 50 MWt level.

  18. Improvement of seismic observation systems in JOYO

    Sumino, Kozo; Suto, Masayoshi; Tanaka, Akihiro

    2013-01-01

    In the experimental fast reactor 'Joyo' in order to perform the seismic observation in and around the building block and ground, SMAC type seismographs had continuously been used for about 38 years. However, this equipment aged, and the 2011 off the Pacific Coast of Tohoku Earthquake on Mach 11, 2011 increased the importance of seismic data of the reactor facilities from the viewpoint of earthquake-proof safety. For these reasons, Joyo updated the system to the seismic observation system reflecting the latest technology/information, while keeping consistency with the observation data of the former seismographs (SMAC type seismograph). This updating improved various problems on the former observation seismographs. In addition, the installation of now observation points in the locations that are important in seismic safety evaluation expanded the data, and further improved the reliability of the seismic observation and evaluation on 'Joyo'. (A.O.)

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

    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)

  20. Development of JOYO operational guidance system for emergency condition

    Takatsuto, Hiroshi; Owada, Toshio; Morimoto, Makoto; Aoki, Hiroshi; Tokita, Mitsuhiko; Terunuma, Seiichi

    1989-01-01

    Operational guidance system in JOYO has been developed for safe and stable plant operations and improvement of operational reliability. JOYCAT (JOYO Consulting and Analysing Tool), one of the JOYO operational guidance systems, supports the plant operator to present the causal alarm and select the suitable guidance manual in anomaly situations using artificial intelligence technology. Verification test of JOYCAT was performed using a JOYO operator-training simulator and on-line operation was started by partially linking to the actual plant in May 1988. As the result, the proper diagnosis function was confirmed in the actual plant. (author)

  1. Development of JOYO plant operation management expert tool

    Michino, Masanobu; Sawada, Makoto [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1995-03-01

    Operation and maintenance support systems for JOYO are being developed in order to keep the stable and safe operation of JOYO and to improve operational reliability of future FBR plants. As one of the systems, an operation support system named JOYPET has been developing and applied. The system supports the plant management works of JOYO which are necessary for much manpower and knowledge of the plant. The plant management of JOYO was able to improve its reliability and reduce manpower by using this system. As a final step, a judgment function based on the accumulated plant management rule of JOYO will be developed and applied. The function judges the plant condition which allows to start the maintenance works or not. (author).

  2. Neutron intensity of fast reactor spent fuel

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  3. Power coefficient anomaly in Joyo, (2)

    Ishikawa, Makoto; Yamashita, Yoshioki; Sasaki, Makoto; Nara, Yoshihiko.

    1981-12-01

    In this report, the presumption about the mechanism having caused the power coefficient anomaly in Joyo during the 75 MW power-raising test in 1979 is described. After the previous report, the new information about the results of the post-irradiation examination and the analysis of the power coefficient of Joyo were able to be obtained. From these information, the mechanism of causing the anomaly was presumed as follows. In 50 MW operation, the fuel burnup reached about 10,000 MWD/ton at the end of second cycle, and produced fission gas was almost retained in fuel pellets. When the power was raised from 50 MW to 75 MW for the first time, the fission gas began to be released when 50 MW was somewhat exceeded. The fission gas release caused the temperature rise and cracking of fuel pellets, and elongated fuel stack length abruptly. These phenomena induced to enlarge the fuel expansion reactivity effect and Doppler reactivity effect, and caused the anomalous behavior of power coefficient. After reaching 75 MW, the fuel stack length did not respond normally to reactor power change, and the magnitude of power coefficient became smaller. The reactivity was lost considerably from the core after the anomaly. (Kako, I.)

  4. Development of JOYO Plant Operation Management Expert Tool (JOYPET)

    Michino, Masanobu; Terano, Toshihiro; Hanawa, Mikio; Aoki, Hiroshi; Okubo, Toshiyuki

    2000-03-01

    The Operation and Maintenance Support Systems for JOYO are being developed, with the aim of ensuring the stable and safe operation of JOYO and improving operational reliability of future FBR plants. Plant Operation Management Expert Tool named JOYPET had been developed as one of the Operation and Maintenance Support Systems, which helps plant operation management. The following functions were developed and applied. (1) Papers management (Plant status management) function for maintenance activities, (2) Isolation management support function for plant operation, (3) Automatically drawing function of plant operation schedule, (4) Isolation judgment function for plant operation. By use this system, the plant management of JOYO was able to improved reliability and reduced manpower. (author)

  5. Plant experience of experimental fast reactor 'Joyo'

    1982-01-01

    The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

  6. Closeout of JOYO-1 Specimen Fabrication Efforts

    ME Petrichek; JL Bump; RF Luther

    2005-01-01

    Fabrication was well under way for the JOYO biaxial creep and tensile specimens when the NR Space program was canceled. Tubes of FS-85, ASTAR-811C, and T-111 for biaxial creep specimens had been drawn at True Tube (Paso Robles, CA), while tubes of Mo-47.5 Re were being drawn at Rhenium Alloys (Cleveland, OH). The Mo-47.5 Re tubes are now approximately 95% complete. Their fabrication and the quantities produced will be documented at a later date. End cap material for FS-85, ASTAR-811C, and T-111 had been swaged at Pittsburgh Materials Technology, Inc. (PMTI) (Large, PA) and machined at Vangura (Clairton, PA). Cutting of tubes, pickling, annealing, and laser engraving were in process at PMTI. Several biaxial creep specimen sets of FS-85, ASTAR-811C, and T-111 had already been sent to Pacific Northwest National Laboratory (PNNL) for weld development. In addition, tensile specimens of FS-85, ASTAR-811C, T-111, and Mo-47.5 Re had been machined at Kin-Tech (North Huntington, PA). Actual machining of the other specimen types had not been initiated. Flowcharts 1-3 detail the major processing steps each piece of material has experienced. A more detailed description of processing will be provided in a separate document [B-MT(SRME)-51]. Table 1 lists the in-process materials and finished specimens. Also included are current metallurgical condition of these materials and specimens. The available chemical analyses for these alloys at various points in the process are provided in Table 2

  7. MISER-I: a computer code for JOYO fuel management

    Yamashita, Yoshioki

    1976-06-01

    A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)

  8. Joyo ATWS test analysis by Mimir-N2

    Yoshida, Akihiro

    2001-03-01

    The study on the passive safety test by using the Experimental Fast Reactor Joyo was performed to demonstrate the inherent safety of fast breeder reactors. An analysis code: Mimir-N2, which has been developed to analyze Joyo plant kinetics, was selected as a standard code for this study. In order to increase the reliability of the calculation, Mimir-N2 code was adjusted based on the data obtained through several plant characteristics tests carried out in Joyo. Throughout an operational data obtained in Joyo, it is supposed that the burn-up dependency observed on the power reactivity coefficient might be coming from the reactivity shift caused by a depression of a thermal expansion of fuel pellet. Based on the relationship between the measured power reactivity coefficient and the core averaged burn-up, the burn-up dependency mentioned above was estimated and introduced to Mimir-N2. As a result, calculated core and plant dynamics during the step reactivity response test, such as the response of the power range neutron monitor and the coolant temperature at the core inlet/outlet, corresponded with the measured value. Especially, it was confirmed that Mimir-N2 can simulate the perturbation caused by the thermal expansion of the core support plate. In addition, Mimir-N2 was modified to be enable to take into account for the core bowing reactivity, which is calculated by the core bowing reactivity analysis system developed for Joyo. The preliminary analysis of the plant dynamics during the ATWS events in MK-III core were carried out by using modified Mimir-N2. As a result, it was confirmed that the core bowing reactivity should not be neglected because it sometimes shows positive feedback characteristics. (author)

  9. Overview of Japanese control rods development program

    Koyama, M.

    1984-01-01

    The Japanese control rods development program was established based on the fast breeder reactor program. Therefore, PNC's efforts have been made mainly for the development of analysis, design and fabrication technologies for ''JOYO'' and ''MONJU'' control rods. Laboratory studies were performed to obtain the information for absorber materials. The design and fabrication of the sealed and vented type control rod pins were completed, and water loop tests and in-sodium tests were carried out. Irradiation behavior of enriched B 4 C pellets with low and high density in DFR was examined. Japan's experimental fast reactor, JOYO, has been operated at the rated power of 50MWt and 75MWt since April 1977 when the MK-I core (breeder core) attained initial criticality. Post irradiation examinations on control rod, removed from the reactor, were carried out and their performance behavior were evaluated. In the MK-II core, a control rods monitoring program has been in investigation. Absorber Materials Irradiation Rigs (AMIR) are scheduled to be loaded and irradiated in the JOYO MK-II core from 1984. (author)

  10. Renewal of JOYO plant operation management expert tool (JOYPET)

    Okawa, Toshikatsu; Aita, Tsuyoshi; Murakami, Takanori; Ito, Hideaki; Aoki; Hiroshi; Oda, Toshihiro

    2004-03-01

    JOYO Plant Operation Management Expert Tool system named JOYPET has been developed with the aim of confirming the stable and safety operation of JOYO and improving operational reliability in future FBR plants. New JOYPET system was designed and manufactured in 2002, and began on operation in 2003, because the former system, which was designed in 1988 and operated from 1991 to 2002, was superannuated, and it was difficult to obtain alternative hardwares and replace parts. The difference between the former system and the later new one was adopted the web-online system to use LAN (Local Area Network) instead of the host and the terminal computer system. Then the new system enabled to take unitary document management for reactor operation, and each person was able to search, refer and make document on line directly. This paper summarized the new JOYPET system design, manufacturing, system constitution and operation actual result. (author)

  11. PNC/DOE Remote Monitoring Project at Japan's Joyo Facility

    Ross, M.; Hashimoto, Yu; Sonnier, C.; Dupree, S.; Ystesund, K.; Hale, W.

    1996-01-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan and the US Department of Energy (DOE) are cooperating on the development of a remote monitoring system for nuclear nonproliferation efforts. This cooperation is part of a broader safeguards agreement between PNC and DOE. A remote monitoring system is being installed in a spent fuel storage area at PNC's experimental reactor facility Joyo in Oarai. The system has been designed by Sandia National Laboratories (SNL) and is closely related to those used in other SNL remote monitoring projects. The Joyo project will particularly study the unique aspects of remote monitoring in contribution to nuclear nonproliferation. The project will also test and evaluate the fundamental design and implementation of the remote monitoring system in its application to regional and international safeguards efficiency. This paper will present a short history of the cooperation, the details of the monitoring system and a general schedule of activities

  12. Simulating the Behaviour of the Fast Reactor Joyo (Draft)

    Juutilainen, Pauli

    2008-01-01

    Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO 2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)

  13. Analysis of control rod worth in experimental fast reactor JOYO

    Arii, Y.; Aoyama, T.; Okimoto, Y.; Yoshida, A.; Mizoo, N.

    1988-01-01

    In JOYO, the measurement of control rod worths have been carried out in the beginning of the each cycle, using both period method and neutron source multiplication method. In this paper, the calculational method of control rod worths in the design stage and the comparison with the design values and measured ones are shown. The reasons that the control rod worths change slightly in each cycle, are also investigated. (author). 13 figs, 12 tabs

  14. Development of FBR technology in the FBR 'Joyo'

    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.)

  15. The analysis of FCA critical experiments and its application to ''JOYO'' nuclear design

    Iijima, S.

    1979-01-01

    A series of extensive mockup experiments in support of Japanese Experimental Fast Reactor, ''JOYO'', were performed at Fast Critical Assembly in JAERI, from February 1970 to March 1972. The present paper describes the results of analysis of these mockup experiments and its application to ''JOYO'' nuclear design. The basic calculational method of the analysis is the same as that employed in ''JOYO'' neutronics calculation, viz., the 6-group diffusion theory using 25-group NAIG Nuclear Set No. 5. Corrections to the base calculations were evaluated by using one-dimensional S 4 transport theory and integral transport theory. The ABBN group constants were also used for the sake of comparison. The most probable values of JOYO neutronics parameters were determined by applying the bias factor (E/C) to the calculated values. The uncertainties of the most probable values were also determined, and they were taken into consideration in the JOYO design

  16. Fast fluence measurement for JOYO irradiation field using niobium dosimeter

    Ito, Chikara

    2004-03-01

    Neutron fluence and spectrum are key parameters in various irradiation tests and material surveillance tests so they need to be evaluated accurately. The reactor dosimetry test has been conducted by the multiple foil activation method, and a niobium dosimeter has been developed for measurement of fast neutron fluence in the experimental fast reactor JOYO. The inelastic scattering reaction of 93 Nb has a low threshold energy, about 30 keV, and the energy distribution of reaction cross section is similar to the displacement cross section for iron. Therefore, a niobium dosimeter is suitable for evaluation of the fast neutron fluence and the displacement per atom for iron. Moreover, a niobium dosimeter is suited to measure neutron fluence in long-term irradiation test because 93 Nb, which is produced by the reaction, has a long half-life (16.4 years). This study established a high precision measurement technique using the niobium reaction rate. The effect of self-absorption was decreased by the solution and evaporation to dryness of niobium dosimeter. The dosimeter weight was precisely measured using the inductively coupled plasma mass spectrometer. This technique was applied to JOYO dosimetry. The fast neutron fluences (E > 0.1 MeV) found by measuring the reaction rate in the niobium dosimeter were compared with the values evaluated using the multiple foil activation method. The ratio of measured fast neutron fluences by means of niobium dosimeter and multiple foil activation method range from 0.97 to 1.03 and agree within the experimental uncertainty. The measurement errors of fast neutron fluence by niobium dosimeter range from 4.5% (fuel region) to 10.1% (in-vessel storage rack). As a result of this study, the high precision measurement of fast neutron fluence by niobium dosimeters was confirmed. The accuracy of fast reactor dosimetry will be improved by application of niobium dosimeters to the irradiation tests in the JOYO MK-III core. (author)

  17. Radiation protection monitoring at the JOYO experimental fast reactor

    Ouchi, S.; Endo, K.; Susaki, T.

    1979-01-01

    This paper describes the radiation protection monitoring programme for the JOYO experimental fast reactor and some of the health physics problems experienced during the low-power nuclear tests. These include: a detailed description of the centralized radiation monitoring system; the methods and results of the individual monitoring systems; the results of operational monitoring for the handling of new plutonium fuel subassemblies; the evaluation of the external radiation dose rate around the primary coolant system; and the results of an experiment on the thermal dependence of some personnel dose meters. (author)

  18. Reactor technology progress report on Joyo, vol. 6

    1982-01-01

    The works of the Technology Section, Fast Experimental Reactor Division, Power Reactor and Nuclear Fuel Development Corp., are roughly divided into core technology, anomaly monitoring techniques, plant technology, purity control techniques and operation planning and management. In this book, the state of activities in the Technology Section, the result of operation of Joyo and the foreign information related to FBRs in the quarter from July to September, 1981, are reported. The operation of Joyo of 75 MW rating No. 5 cycle was finished on August 9, and after fuel handling and FFDL test, the operation of special test cycle was carried out in September. In this quarter, main report papers were one N-report and 108 memos. The examination of the preliminary analysis and the plan for shifting to the MK-2 core and the performance test, and the planning of the core construction for the operation from No. 1 to No. 3 cycle with the MK-2 core and the analysis of its characteristics were carried out. The revision of the long term plan of the Technology Section was started in July, and the first draft was completed in September. The compilation of the general report on the MK-1 core was started in July. Three meetings for technical discussion within the Division were held. (Kako, I.)

  19. JOYO operation support system 'JOYCAT' based on intelligent alarm handling

    Tamaoki, Tetsuo; Yamamoto, Hiroki; Sato, Masuo; Yoshida, Megumu; Kaneko, Tomoko; Terunuma, Seiichi; Takatsuto, Hiroshi; Morimoto, Makoto.

    1992-01-01

    An operation support system for the experimental fast reactor 'JOYO' was developed based on an intelligent alarm-handling. A specific feature of this system, called JOYCAT (JOYO Consulting and Analyzing Tool), is in its sequential processing structure that a uniform treatment by using design knowledge base is firstly applied for all activated alarms, and an exceptional treatment by using heuristic knowledge base is then applied only for the former results. This enables us to achieve real-time and flexible alarm-handling. The first alarm-handling determines the candidates of causal alarms, important alarms with which the operator should firstly cope, through identifying the cause-consequence relations among alarms based on the design knowledge base in which importance and activating conditions are described for each of 640 alarms in a frame format. The second alarm-handling makes the final judgement with the candidates by using the heuristic knowledge base described as production rules. Then, operation manuals concerning the most important alarms are displayed to operators. JOYCAT has been in commission since September of 1990, after a wide scope of validation tests by using an on-site full-scope training simulator. (author)

  20. Development of wire wrapping technology for FBR fuel pin

    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)

  1. Production of analysis code for 'JOYO' dosimetry experiment

    Sasaki, Makoto; Nakazawa, Masaharu.

    1981-01-01

    As part of the measurement and analysis plan for the Dosimetry Experiment at the ''JOYO'' experimental fast reactor, neutron flux spectra analysis is performed using the NEUPAC (Neutron Unfolding Code Package) computer program. The code calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils. The NEUPAC code is based on the J1-type unfolding method, and the estimated neutron flux spectra is obtained as its solution. The program is able to determine the integral quantities and their sensitivities, together with an error estimate of the unfolded spectra and integral quantities. The code also performs a chi-square test of the input/output data, and contains many options for the calculational routines. This report presents the analytic theory, the program algorithms, and a description of the functions and use of the NEUPAC code. (author)

  2. Operational experience of the fuel cleaning facility of Joyo

    Mukaibo, R.; Matsuno, Y.; Sato, I.; Yoneda, Y.; Ito, H.

    1978-01-01

    Spent fuel assemblies in 'Joyo', after they are taken out of the core, are taken to the Fuel Cleaning Facility in the reactor service building and sodium removal is done. The cleaning process is done by cooling the assembly with argon gas, steam charging and rinsing by demineralized water. Deposited sodium was 50 ∼ 60 g per assembly. The sodium and steam reaction takes about 15 minutes to end and the total time the fuel is placed in the pot is about an hour. The total number of assemblies cleaned in the facility was 95 as of November 1977. In this report the operational experience together with discussions of future improvements are given. (author)

  3. Integrated leak rate test results of JOYO reactor containment vessel

    Tamura, M.; Endo, J.

    1982-02-01

    Integrated leak rate tests of JOYO after the reactor coolant system had been filled with sodium have been performed two times since 1978 (February 1978 and December 1979). The tests were conducted with the in-containment sodium systems, primary argon cover gas system and air conditioning systems operating. Both the absolute pressure method and the reference chamber method were employed during the test. The results of both tests confirmed the functioning of the containment vessel, and leak rate limits were satisfied. In Addition, the adequancy of the test instrumentation system and the test method was demonstrated. Finally the plant conditions required to maintain reasonable accuracy for the leak rate testing of LMFBR were established. In this paper, the test conditions and the test results are described. (author)

  4. Operational experience of the fuel cleaning facility of Joyo

    Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Ito, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

    1978-08-01

    Spent fuel assemblies in 'Joyo', after they are taken out of the core, are taken to the Fuel Cleaning Facility in the reactor service building and sodium removal is done. The cleaning process is done by cooling the assembly with argon gas, steam charging and rinsing by demineralized water. Deposited sodium was 50 {approx} 60 g per assembly. The sodium and steam reaction takes about 15 minutes to end and the total time the fuel is placed in the pot is about an hour. The total number of assemblies cleaned in the facility was 95 as of November 1977. In this report the operational experience together with discussions of future improvements are given. (author)

  5. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  6. Reactor noise analysis of experimental fast reactor 'JOYO'

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  7. Analysis of contamination conditions of the Joyo Waste Treatment Facility

    Yoshizawa, S.; Ishijima, N.; Tanimoto, K.

    1999-08-01

    Decontamination methods have been studied for decommissioning of Joyo Waste Treatment Facility whose operation has been stopped in 1994. In this study, we analyzed samples of its system piping, whose dose rate was relatively low, to determine conditions of contamination. We also study appropriate decontamination methods for them. Results are as follows. 1. The inner surfaces of piping were covered with a very thin clad that was less than 1 micrometer in thickness and had many vacancies, looked like particle detachment, which were about 20 micrometers in depth. Something like corrosion product was observed near the surface and it was 440 micrometers in depth. 2. Radioactive contamination was considered to settle on a lower part of the piping and to be buried in the clad. A kind of dominant contamination nuclide was 60 Co. 3. Hot nitric acid process will be suitable for system decontamination to reduce dose rate before dismantling. But its feasibility tests are indispensable using samples of main system components that have high dose rate. Rubber lining tanks requires another methods because of its difficulty of decontamination. 4. Analyses and decontamination tests using main system are required to decide through decontamination methods according to the clearance level. (author)

  8. Operation experiences of JOYO fuel failure detection system

    Tamura, Seiji; Hikichi, Takayoshi; Rindo, Hiroshi.

    1982-01-01

    Monitoring of fuel failure in the experimental fast reactor JOYO is provided by two different methods, which are cover gas monitoring (FFDCGM) by means of a precipitator, and delayed neutron monitoring (FFDDNM) by means of neutron detectors. The interpretation of signals which were obtained during the reactor operation for performance testings, was performed. The countrate of the CGM is approximately 120 cps at 75MW operation, whose sources are due to Ne 23 , Ar 41 , and Na 24 . And the countrate of the DNM is approximately 2300 cps at 75MW operation which is mainly due to leakage neutron from the core. With those background of the systems, alarm level for monitoring was set at several times of each background level. The reactor has been operated for 5 years, the burn-up of the fuel is 40,000 MWD/T at the most. No trace of any fuel failure has been observed. The fact is also proven by the results of cover gas and sodium sampling analysis. In order to evaluate sensitivity of the FFD systems, a preliminary simulation study has been performed. According to the results, a signal level against one pin failure of 0.5 mm 2 hole may exceed the alarm level of the FFDCGM system. (author)

  9. Processing of Refractory Metal Alloys for JOYO Irradiations

    RF Luther; ME Petrichek

    2006-01-01

    This is a summary of the refractory metal processing experienced by candidate Prometheus materiats as they were fabricated into specimens destined for testing within the JOYO test reactor, ex-reactor testing at Oak Ridge National Laboratory (ORNL), or testing within the NRPCT. The processing is described for each alloy from the point of inception to the point where processing was terminated due to the cancellation of Naval Reactor's involvement in the Prometheus Project. The alloys included three tantalum-base alloys (T-111, Ta-10W, and ASTAR-811C), a niobium-base alloy, (FS-85), and two molybdenum-rhenium alloys, one containing 44.5 w/o rhenium, and the other 47.5 w/o rhenium. Each of these alloys was either a primary candidate or back-up candidate for cladding and structural applications within the space reactor. Their production was intended to serve as a forerunner for large scale production ingots that were to be procured from commercial refractory metal vendors such as Wah Chang

  10. Calculational and experimental experience on core management of experimental fast reactor 'JOYO'

    Yoshida, A.; Arii, Y.; Shono, A.; Suzuki, S.; Kinjo, K.

    1992-01-01

    For the core management of JOYO Mark-II, many core characteristics have been calculated with the core management code system 'MAGI', and measurements have also been carried out at each duty operation cycle. From the evaluation of these results, the characteristics of core parameters such as criticality, reactivity coefficients, and control rod worth can be predicted accurately as followings; excess reactivity: ± 0.1% Δk/k, outlet temperature of subassembly: ±10degC, fuel burn-up: ±5%, control rod worth: ±5%. As a result, we can not only get steady operation of JOYO but also perform various irradiation tests with satisfied conditions. This paper presents experience obtained until now through twenty three duty cycle operations of Mark-II core in JOYO. (author)

  11. JOYO modification program for demonstration tests of FBR innovative technology development

    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

  12. Irradiation tests report of the 32nd cycle in 'JOYO'

    1998-09-01

    This report summarizes the operating and irradiation data of the experimental reactor 'JOYO' 32nd cycle, and estimates the 33rd cycle irradiation condition. Irradiation tests in the 31st cycle are as follows: (1) B-type irradiation rig (B9). (a) High burn up performance tests of MONJU' fuel pins, advanced austenitic steel cladding fuel pins, large diameter fuel pins, ferrite steel cladding fuel pins (in collaboration with the USA) and large diameter annular pellet fuel pins. (b) Mixed carbide and nitride fuel pins irradiation tests (in collaboration with JAERI). (2) C-type irradiation rig (C4F). (a) High burn up performance test of advanced austenitic steel cladding fuel pins (in collaboration with France). (3) C-type irradiation rig (C6D). (a) Large diameter fuel pins irradiation test. (4) Absorber Materials Irradiation Rig (AMIR-6). (a) Run to absorber pin's cladding breach. (5) Absorber Materials Irradiation Rig (AMIR-8). (a) High-temperature shroud and Na-bond elements tests. (6) Core Materials Irradiation Rig (CMIR-5-1). (a) Core materials irradiation tests. (7) Structure Materials Irradiation Rigs (SMIR). (a) Material irradiation tests (in collaboration with universities). (b) Surveillance back up tests for MONJU'. (8) MAterial testing RIg with temperature COntrol (MARICO-1). (a) Material irradiation tests (in collaboration with universities), (b) Creep rupture tests of the core materials for the demonstration reactor. (9) Upper core structure irradiation Plug Rig (UPR-1-5). (a) Upper core neutron spectrum effect and accelerated irradiation effect. The maximum burn-up driver assembly 'PFD503' reached 65,600 MWd/t (pin average). (author)

  13. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    G. Borges

    2006-01-01

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus

  14. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    G. Borges

    2006-01-31

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.

  15. Education and training for operators using a full scope simulator and an its upgrading program in JOYO

    Sawada, Makoto; Terano, Toshihiro; Hunaki, Isao

    1996-01-01

    A JOYO full scope operator training simulator installed in 1983, is being used with high average unit availability factor of more than 70% per annum. The education and training for the operators using it has been greatly contributing to safety operation of the experimental fast reactor JOYO. The simulator mainly consisting of five control panels, a computer system having two computers and an instructor's console, is able to simulate the plant behaviors and the sequential processes with real time under normal or anomaly conditions. Now, according as the JOYO MK-ILL project which enhances the irradiation capability of JOYO, an upgrading program of the simulator is proceeding with the aim of advancing its efficient usage by improving the training function and the analytical accuracy of the simulator. (author)

  16. A study of reactor neutrino monitoring at the experimental fast reactor JOYO

    Furuta, H.; Fukuda, Y.; Hara, T.; Haruna, T.; Ishihara, N.; Ishitsuka, M.; Ito, C.; Katsumata, M.; Kawasaki, T.; Konno, T.; Kuze, M.; Maeda, J.; Matsubara, T.; Miyata, H.; Nagasaka, Y.; Nitta, K.; Sakamoto, Y.; Suekane, F.; Sumiyoshi, T.; Tabata, H.

    2012-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3 m from the JOYO reactor core of 140 MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11±1.24(stat.)±0.46(syst.) events/day. Although the statistical significance of the measurement was not enough, backgrounds in such a compact detector at the ground level were studied in detail and MC simulations were found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  17. The off-line computation system for supervising performance of JOYO: JOYPAC system, 1

    Katsuragi, Satoru; Inoue, Teruji; Shimizu, Akinao; Yoshino, Fujio; Suzuki, Masao.

    1976-10-01

    A code system JOYPAC for monitoring the operation of the fast experimental reactor JOYO has been developed. This is an off-line code system designed for use in making calculation of the nuclear and thermohydraulic characteristics of the reactor core and also to make computation of the history of core irradiation after reactor operation. The use of the code system makes it possible to calculate the various core characteristics with a high degree of accuracy by simplified procedure for the diverse operation patterns of JOYO to confirm its safety. It also enables the details of the history of irradiation of the core to be obtained quickly and accurately after reactor operation. The above include all the operation data and in-pile characteristics that are required for the irradiation test. Furthermore, it is also possible to provide the data for the on-line computer system of JOYO and the data for nuclear material accountability. The code system consists of the detailed subsystem and the simplified subsystem. The former is used for obtaining the nuclear and thermohydraulic characteristics of the core by use of a detailed calculation model such as three-dimensional hexagonal lattice, for instance, in order to back up the simplified subsystem. On the other hand, the latter is designed to obtain the various core characteristics by use of simple extrapolation and interpolation methods, whose conception is based on the great deal of information obtained by the design calculation of JOYO and the many parameter surveys. The system is used for the normal cycle operation. (J.P.N.)

  18. Specialist committee's review reports for experimental fast reactor JOYO' MK-III performance tests

    Yamashita, Kiyonobu; Okubo, Toshiyuki; Kamide, Hideki

    2004-02-01

    Performance tests (startup-physics tests and power elevation tests) were planed for experimental fast reactor 'JOYO' MK-III where irradiation performances were upgraded by power increase from 100 to 140 MW. The reactor safety committee of O-arai Engineering Center has established a specialist committee for 'JOYO' MK-III Performance Tests at the first meeting of 2003 on 23th. April 2003, to accomplish the tests successfully. Subjects of the specialist committee were reviews of following items covering a wide range. 1) Contents of modification works. 2) Reflections of functional test results to the plant and facilities. 3) Reflections of safety rule modification to instruction and manual for operation. 4) Quality assurances and pre-calculation for performance test. 5) Inspection plan and its results. 6) Adequacy of performance test plan. 7) Confirmation of performance test results. Before test-starts, the specialist committee has confirmed by reviewing the items from 1) to 6) based on explanations and documents of the Division of Experimental Reactor, that the test plan and pre-inspections are adequate. After the tests, the specialist committee had confirmed by reviewing the item 7) in the same way, that the each test result satisfies the corresponding criterion. The specialist committee has concluded from these review's results before and after the tests that the 'JOYO' MK-III Performance Tests were carried out appropriately. Besides, the first criticality of the JOYO MK-III was achieved on 2nd. July 2003, and the continuous full power operation was carried on 20th. Nov. 2003. Finally, all performance tests were completed by the pass of the last governmental pre-serviced inspection (dose rate measurement during the shut down condition). (author)

  19. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

  20. Inhibitory effects of KN-93, an inhibitor of Ca2+ calmodulin-dependent protein kinase II, on light-regulated root gravitropism in maize

    Feldman, L. J.; Hidaka, H.

    1993-01-01

    Light is essential for root gravitropism in Zea mays L., cultivar Merit. It is hypothesized that calcium mediates this light-regulated response. KN-93, an inhibitor of calcium/calmodulin kinase II (CaMK II), inhibits light-regulated root gravitropism but does not affect light perception. We hypothesize that CaMK II, or a homologue, operates late in the light/gravity signal transduction chain. Here we provide evidence suggesting a possible physiological involvement of CaMK II in root gravitropism in plants.

  1. Irradiation test of fuel containing minor actinides in the experimental fast reactor Joyo

    Soga, Tomonori; Sekine, Takashi; Wootan, David; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    2007-01-01

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP, accounting for both prompt and delayed heating components, and then adjusted using E/C for 10 B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO 2-x or AmO 2-x in the (U, Pu)O 2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel. (author)

  2. Experimental fast reactor JOYO MK-III functional test. Primary auxiliary cooling system test

    Karube, Koji; Akagi, Shinji; Terano, Toshihiro; Onuki, Osamu; Ito, Hideaki; Aoki, Hiroshi; Odo, Toshihiro

    2004-03-01

    This paper describes the results of primary auxiliary cooling system, which were done as a part of JOYO MK-III function test. The aim of the tests was to confirm the operational performance of primary auxiliary EMP and the protection system including siphon breaker of primary auxiliary cooling system. The items of the tests were: (Test No.): (Test item). 1) SKS-117: EMP start up test. 2) SKS-118-1: EMP start up test when pony motor running. 3) SKS-121: Function test of siphon breaker. The results of the tests satisfied the required performance, and demonstrated successful operation of primary auxiliary cooling system. (author)

  3. 2010 Great Lakes Restoration Initiative Bathymetric Lidar: Lake Superior

    National Oceanic and Atmospheric Administration, Department of Commerce — The data contained in this file contain hydrographic and topographic data collected by the Fugro LADS Mk II system along the Lake Superior coast of Minnessota,...

  4. Evaluation of flow-induced vibration of thermometer well for JOYO

    Isozaki, Kazunori; Tomita, Naoki

    1997-05-01

    Sodium leak accident of MONJU was caused high cycles fatigue damage of thermometer well by flow-induced vibration. It was due to the symmetric vortex shedding which was occurred rear flow of thermometer well. So, Thermometer wells installed in primary and secondary heat transport systems of JOYO were evaluated of flow-induced vibration. Evaluation of flow-induced vibration of thermometer well was done checking of flow-induced vibration base on authorized design report for JOYO, evaluation of summary flow-induced vibration by natural frequency of thermometer well in sodium as cantilever models, and evaluation based on small velocity rule of ASME Code Section III Appendix N-1300. By this result, thermometer wells (12B piping of secondary cooling system) were not satisfied requirement to avoid flow-induced vibration by small velocity rule. Therefore, Detailed vibration characteristic analysis, water flow-induced vibration test, dumping test and evaluation of structural integrity were carried out. These results, vibration amplitude of well on the tip was 0.13 mm (vibration non-dimensional amplitude of 0.015) and peak stress of 2.9 kg/mm 2 is occurred. Thermometer wells (12B piping of secondary cooling system) which occurred peak stress by flow vibration was confirmed enough to satisfy 5.3 kg/mm 2 of design fatigue limit. (author)

  5. Construction of fast experimental reactor 'Joyo' from start of construction to criticality

    Sakata, Hajime

    1977-01-01

    The fast experimental reactor ''Joyo'' is a sodium-cooled, fast neutron reactor using mixed oxide of uranium and plutonium, the first in Japan. The purposes of its construction are to experience and solve the various technical problems expected in the constructions of the prototype reactor ''Monju'' and future practical reactors, and to use as the irradiation facility for developing the fuel and material for fast breeder reactors in Japan after the completion. The construction finished by the end of 1974, and the synthetic functional test was carried out for about two years thereafter. The whole installation was handed over to PNC on March 8, 1977. The reactor attained the criticality on April 24, 1977. The outline of the construction works is described. ''Guidance to the structural design of sodium machinery for Joyo'' was compiled, and the analysis was made according to it. Moreover, various inspection standards regarding welding, electrical machinery, fuel and others were made. The revision of the design for improving the safety and performance was made during the construction at all times. The synthetic functional test was carried out for about two years on 266 items, and subsequently, the criticality test was completed satisfactorily. (Kako, I.)

  6. Summary report of the experimental fast reactor JOYO MK-III performance test

    Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro

    2004-03-01

    An upgrading project (MK-III project) was started to improve the irradiation capability of the experimental fast reactor JOYO. In this project, core replacement and increase of the reactor thermal power by the factor 1.4 were necessary for increasing the maximum fast neutron flux by the factor 1.3 and doubling the capacity for irradiation rigs. The modification of the cooling system that included the replacement of the main intermediate heat exchangers and the dump heat exchangers was completed in September 2000. After a series of system function tests, the performance test, of which objective is to fully characterize the upgraded core and heat transfer system, was started in June 2003. Twenty eight tests were selected and carried out as performance test, in order to confirm that the whole plant satisfy the design criteria and have sufficient characteristics (data necessary for safe and steady operation, core management, reactor control and monitoring) as an irradiation bed. After attaining the initial criticality of the core on 2nd July 2003, core characteristics (the excess reactivity, the isotherm temperature reactivity coefficient, the power reactivity coefficient and so on), plant characteristics (the plant heat balance, the adjustment of the temperature control system, the plant behavior at transient), shielding characteristics (dose rate distribution). As the result, it was confirmed that all the criteria regulated was satisfied and the core and plant have sufficient margins for full power operation, which was increased by the factor 1.4. Especially, nuclear analysis accuracy was verified by comparing the calculation with measured core characteristics of the initial core which consists of fifty five fresh fuel subassemblies. The operational data which is supposed to be useful for developing in-core anomaly detection system were also obtained. The operation manual and training simulator and design of next reactor development were revised based on the results

  7. Periodic safety review of the experimental fast reactor JOYO. Review of the activity for safety

    Maeda, Yukimoto; Kashimura, Youichi; Suzuki, Toshiaki; Isozaki, Kazunori; Hoshiba, Hideaki; Kitamura, Ryoichi; Nakano, Tomoyuki; Takamatsu, Misao; Sekine, Takashi

    2005-02-01

    Periodic safety review (Review of the activity for safety) which consisted of 'Comprehensive evaluation of operation experience' and Incorporation of the latest technical knowledge' was carried out up to January 2005. 1. Comprehensive evaluation of operation experience. It was confirmed that the effectual activities for safety through the operation of JOYO were carried out in terms of (1) Operation management, (2) Maintenance management, (3) Fuel management, (4) Radiation management, (5) Radioactive waste management, (6) Emergency planning and (7) Feedback of incidents and failures. 2. Reflection of the latest technical knowledge. It was confirmed that the latest technical knowledge including regulation and guide line established by Nuclear Safety Commission of Japan until March 31st. 2003 were properly reflected in impressing the safety of the reactor. As a result, it was evaluated that the activity for safety was carried out effectually, and no additional measure was identified continual safe operation of the reactor. (author)

  8. Aspects of 238Pu production in the experimental fast reactor JOYO

    Osaka, Masahiko; Koyama, Shin-ichi; Tanaka, Kenya; Itoh, Masahiko; Saito, Masaki

    2005-01-01

    Experimental determination of 238 Pu in 237 Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of 238 Pu production from 237 Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the 238 Pu isotopic ratio. 238 Pu production amount in the irradiated 237 Np samples was determined by a radioanalytical technique. Aspects of 238 Pu production were examined on the basis of the present radioanalysis. The 238 Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu

  9. Study on radioactive corrosion products behaviour in primary circuits of JOYO

    Iizawa, Katsuyuki; Suzuki, Soju; Tamura, Masaaki; Seki, Seiichi; Hikichi, Takayoshi

    1987-01-01

    Radioactive CP deposition and distribution, and the resulting radiation fields along the JOYO primary circuit piping have been measured. The measurement results have been compared with calculations for estimating radioactive CP behaviour and the resulting radiation fields in an LMFBR primary circuit using a computer code which is named PSYCHE. The deposited radioactivity of CPs calculated by using PSYCHE agreed well with the measured results within a factor of 0.5-2. The gamma dose rate distribution calculated from the PSYCHE results reproduced measured values within a factor of 0.6-2 over the piping system, using the JOANDARC modification of the QAD-CG code. Using these verified codes, a prediction of radiation levels for future plant operation, and an evaluation of methods for the reduction of radioactive CPs have been conducted. (orig./DG)

  10. Measurement and evaluation of radioactive corrosion product behaviour in primary sodium circuits of JOYO

    Ito, K.; Iizawa, K.; Takahashi, K.; Zulquarnain, M.A.; Suzuki, S.; Kinjo, K.

    1992-01-01

    In the experimental fast reactor JOYO, the radioactive corrosion product (CP) measurement has been conducted in the primary sodium circuits during each annual inspection. The measured data has been analyzed by the computer code 'PSYCHE', which has been developed by PNC. Main results obtained from the measurements and/or calculations are as follows; (1) The dominant CP nuclide is 54 Mn followed by 60 Co and 58 Co. (2) Average surface gamma dose rate around the primary piping system at the 8th annual inspection is 0.96 mSv/h. The increasing rate of this value is 0.25 (mSv/h)/EFPY. (3) The calculated deposition densities of 54 Mn and 60 Co agree with measured ones within factor of 0.7 ∼ 1.7. (author)

  11. Measurement and calculation of radiation sources in the primary cooling system of JOYO

    Suzuki, S.; Iizawa, K.; Ohtani, N.; Kobayashi, T.; Horie, J.; Handa, H.

    1987-01-01

    Production and transfer of radiation sources in the primary cooling system are important consideration in the LMFBR plant from the viewpoint of radiation protection and shielding design. These items were evaluated with calculations and/or measurements in the Japanese experimental fast reactor JOYO. In this study, calculations were made with the DOT3.5 0 two-dimensional discrete ordinate transport code to determine the neutron flux and production rate distributions of radiation sources in the reactor vessel. Using the DOT results, the behavior in primary coolant sodium of the CP (radioactive corrosion products) which were released from the reactor structural material was also calculationally analyzed with the PSYCHE code developed by PNC. These analytical results were compared with the measured results to get the verification of analysis methods and to estimate the accuracy of calculations

  12. Validation of intermediate heat and decay heat exchanger model in MARS-LMR with STELLA-1 and JOYO tests

    Choi, Chiwoong; Ha, Kwiseok; Hong, Jonggan; Yeom, Sujin; Eoh, Jaehyuk [Sodium-cooled Fast Reactor Design Division, Korea Atomic Energy Research Institute (KAERI), 989-111, Daedeok-Daero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Jeong, Hae-yong, E-mail: hyjeong@sejong.ac.kr [Department of Nuclear Engineering, Sejong University, 209 Neungdong-ro, Gwangjin-gu, Seoul 143-747 (Korea, Republic of)

    2016-11-15

    Highlights: • The capability of the MARS-LMR for heat transfer through IHX and DHX is evaluated. • Prediction of heat transfer through IHXs and DHXs is essential in the SFR analysis. • Data obtained from the STELLA-1 and the JOYO test are analyzed with the MARS-LMR. • MARS-LMR adopts the Aoki’s correlation for tube side and Graber-Rieger’s for shell. • The performance of the basic models and other available correlations is evaluated. • The current models in MARS-LMR show best prediction for JOYO and STELLA-1 data. - Abstract: The MARS-LMR code has been developed by the Korea Atomic Energy Research Institute (KAERI) to analyze transients in a pool-type sodium-cooled fast reactor (SFR). Currently, KAERI is developing a prototype Gen-IV SFR (PGSFR) with metallic fuel. The decay heat exchangers (DHXs) and the intermediate heat exchangers (IHXs) were designed as a sodium-sodium counter-flow tube bundle type for decay heat removal system (DHRS) and intermediate heat transport system (IHTS), respectively. The IHX and DHX are important components for a heat removal function under normal and accident conditions, respectively. Therefore, sodium heat transfer models for the DHX and IHX heat exchangers were added in MARS-LMR. In order to validate the newly added heat transfer model, experimental data were obtained from the JOYO and STELLA-1 facilities were analyzed. JOYO has two different types of IHXs: type-A (co-axial circular arrangement) and type-B (triangular arrangement). For the code validation, 38 and 39 data points for type A and type B were selected, respectively. A DHX performance test was conducted in STELLA-1, which is the test facility for heat exchangers and primary pump in the PGSFR. The DHX test in STELLA-1 provided eight data points for a code validation. Ten nodes are used in the heat transfer region is used, based on the verification test for the heat transfer models. RMS errors for JOYO IHX type A and type B of 19.1% and 4.3% are obtained

  13. Development of the automatic control rod operation system for JOYO. Verification of automatic control rod operation guide system

    Terakado, Tsuguo; Suzuki, Shinya; Kawai, Masashi; Aoki, Hiroshi; Ohkubo, Toshiyuki

    1999-10-01

    The automatic control rod operation system was developed to control the JOYO reactor power automatically in all operation modes(critical approach, cooling system heat up, power ascent, power descent), development began in 1989. Prior to applying the system, verification tests of the automatic control rod operation guide system was conducted during 32nd duty cycles of JOYO' from Dec. 1997 to Feb. 1998. The automatic control rod operation guide system consists of the control rod operation guide function and the plant operation guide function. The control rod operation guide function provides information on control rod movement and position, while the plant operation guide function provide guidance for plant operations corresponding to reactor power changes(power ascent or power descent). Control rod insertion or withdrawing are predicted by fuzzy algorithms. (J.P.N.)

  14. Development of observation techniques in reactor vessel of experimental fast reactor Joyo

    Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto

    2010-01-01

    In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1 mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8 mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to

  15. Study on in-vessel ISI for JOYO. Ultrasound propagation characteristic in the core support plate

    Ariyoshi, Masahiko; Ara, Kuniaki; Hirabayashi, Masaru

    2005-03-01

    The report describes the feasibility study on the in-vessel inspection technique to be applied for the experimental fast reactor JOYO. The object of this examination is to confirm the integration of reactor structure under sodium environment by an immediate means. The core support plate which is an important structure supports the weight of the core assembly is selected to an object of the inspection. In the examination until last year, the core support plate inspection equipment concept which combined ultrasound sensor with manipulator was constructed. In this concept, the ultrasound sensor is accessed to a low-pressure plenum sidewall and integrity of the core support plate weld is inspected. In this study, the ultrasound propagation behavior was examined to confirm the range where the core support plate by this concept was able to be inspected. The outline result is shown follows. (1) Only the transverse wave can be generated in the structure material by reflecting the incidence longitudinal wave from the sensor in the wedge. The use of this transverse wave is effective in the core support plate inspection. (2) Because the attenuation of the ultrasound wave depends on the distance, the sensor is made to approach from the fuel rack in the reactor vessel about two places in the upper part of the core support plate weld far from low-pressure plenum. (3) It is necessary to evaluate the permeability of the ultrasound wave by the mock-up examination in consideration of a peculiar attenuation of the structural material, the reflectivity from defect, etc. (4) In the core support plate inspection of phenix reactor, a weld about 4m away from the sensor position is inspected by using the Lamb wave. In this inspection, because it was generated to echo according to the geometrical shape of the structure material, the evaluation method by the analysis to identify the echo from the defect was constructed, and it was verified by the mock-up examination. It is preferable that

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

    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.)

  17. Measurement of burnup in FBR MOX fuel irradiated to high burnup

    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)

  18. Earthquake response characteristics of large structure 'JOYO' deeply embedded in quaternary ground, (3)

    Yajima, Hiroshi; Sawada, Yoshihiro; Hanada, Kazutake; Sawada, Makoto.

    1987-01-01

    In order to examine aseismicity of embedded structure and to clarify embedment effect, earthquake observations of the large structure 'JOYO' are carried out which is deeply embedded in quaternary ground, and the results are summarized as follows. (1) Amplification factors of horizontal component in ground surface is about 3 to 4 times against the bedrock. Contrastively on the structure, any amplification is not observed at the underground portion, however, little amplification exists at the ground portion of structure. (2) Transfer function of structure has several predominant peaks at frequencies of 4.3 Hz and 8.0 Hz which are well coincided with values obtained from force excitation tests. It is shown that transfer function between basement and ground surface is similar to that between ground of same level to basement and ground surface, suggesting the behavior of basement to be able to estimate by these under ground earthquake motion. (3) According to earthquake motion analysis using S-R models, without regard to consider or not the side ground stiffness, the calculated response values do not so much differ in each model and mostly correspond with observation data, provided that the underground earthquake motion at same level to basement is used as a input wave. Consequently, the behavior of these deeply embedded structure is subject to setting method of input wave rather than modeling method, and it is very useful in design that the most simple model without side ground stiffness can roughly represent the embedment effect. (author)

  19. Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-08-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor "JOYO" to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.

  20. Change in physical properties of high density isotropic graphites irradiated in the ''JOYO'' fast reactor

    Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

    1995-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor ''JOYO'' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597 C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (orig.)

  1. Irradiation behavior evaluation of oxide dispersion strengthened ferritic steel cladding tubes irradiated in JOYO

    Yamashita, Shinichiro, E-mail: yamashita.shinichiro@jaea.go.jp; Yano, Yasuhide; Ohtsuka, Satoshi; Yoshitake, Tsunemitsu; Kaito, Takeji; Koyama, Shin-ichi; Tanaka, Kenya

    2013-11-15

    Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel. Two types of ODS (9Cr–ODS{sub F}/M, 12Cr–ODS{sub F}) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained. Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS{sub F}/M steel whereas no distinct temperature dependence for 12Cr–ODS{sub F} steel. Also, there was no significant change in tensile strengths after irradiation below 923 K, but those above 1023 K up to 6.6 × 10{sup 26} n/m{sup 2} (E > 0.1 MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test.

  2. Sodium removal from the grapples of the fuel handling facility of Joyo

    Mukaibo, R; Matsuno, Y; Sato, I; Yoneda, Y; Sato, H [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

    1978-08-01

    Sodium removal from the grapples of the fuel handling facility of 'JOYO' is done in alcohol. The operations of the cleaning facility started as the functional tests of the fuel handling facility began. Since then, criticality test and low power tests had been done and during this period, sodium removal from the grapples, after a certain amount of time in use, were done. In order to lessen the time for the cleaning process for the grapples of the machines inside the containment vessel, demineralized water concentration in the alcohol was gained to as much as 10% and good results were obtained. On the other hand, there were very small amounts of sodium on the grapples of the machine used outside the containment vessel and direct charging of demineralized water into the cleaning pot was done experimentally, also with good results. In this report, the sodium removal experience of the grapples before power up tests and some remarks on the improvements of the facility for the future are presented. (author)

  3. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  4. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

    2014-01-01

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  5. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    Currently, the lifetime of control rods in JOYO is limited by Absorber-Cladding Mechanical Interaction (ACMI) due to swelling of B{sub 4}C(boron carbide) pellets accelerated by relocation of pellet fragments. A sodium bonded type control rod was developed which improves the thermal conductivity by means of charging sodium into the gap between B{sub 4}C and cladding and by utilizing a shroud which wraps the pellet fragments in a thin tube. This new design will be able to enlarge the gap between B{sub 4}C and cladding, without heating B{sub 4}C or fragment relocation, thus extending the life of the control rod. The sodium bonded type will be fabricated as the ninth reload control rods in JOYO. (1) The specification of a sodium bonded type control rod was determined with the wide gap between B{sub 4}C and cladding. In the design simulation, main component temperature were below the maximum limit. And the local heating by helium bubble generated from B{sub 4}C in the sodium gap, was not a serious problem in the analysis which was considered. (2) A structural design for the sodium entrance into the pin was determined. A formula was developed which the limit for sodium charging given physical dimension of the structure and sodium property. Result from sodium out-pile experiments validated the theoretical formula. (3) The analysis of ACMI indicated a lifetime extension of the sodium bonded type by 4.6% in comparison with lifetime of the helium bonded type of 1.6%. This is due to the boron10 burn-up rate being three times higher in the sodium bonded type than in the helium bonded type. To achieve a target burn-up 10% in the future, it will be necessary to modify design based on irradiation data which will be obtained by practical use of the sodium bonded control rods in JOYO. (4) The effects due to Absorber-Cladding Chemical Interaction (ACCI) were reduced by controlling the cladding temperature and chromium coating to the cladding's inner surface. It was confirmed

  6. JOYO MK-III modification work on heat transport system. Working plan and plant control

    Isozaki, K.; Ichige, S.; Ohshima, J.

    2002-07-01

    The MK-III project to improve the irradiation capability of the experimental fast reactor JOYO have been in underway since 1987. The increase of fast neutron flux and the enlargement of that field increase the reactor thermal rate from 100 MWt to 140 MWt. To increase cooling capacity of heat transport system, intermediate heat exchangers (IHXs), dump heat exchangers (DHXs), piping connecting to IHXs and DHXs, main motors on primary and secondary main circulation pumps were replaced. The replacement of these large components was carried out under following hard conditions. 1) Limitation of work space, 2) Fuel subassembly and molten sodium in the reactor vessel, 3) high radiation circumstances for primary cooling system, 4) treatment of radioactive sodium (radioactive sodium and corrosion product such as 60 Co, 54 Mn). There are little experiences of this kind of work in the world. Therefore the organization, working plan and safety management points were carefully examined and established, based on the previous experience of JOYO operation and maintenance, research and development results of safety treatment of sodium, experience of previous work on sodium facilities. Followings results were obtained and effectiveness was confirmed in the work. (1) Development of most suitable working plan derived from elements and full size mock up experiments, reduction of exposure time by workers training, reduction of radiation dose by installation of temporal radiation shielding were useful to reduce radiation dose. The usage of seal bag was useful to prevent the contamination spreading over. (2) The usage of seal bag, oxygen concentration monitoring in the seal bag, nitrogen concentration monitoring in the cooling system cover gas, low pressure control of cover gas were useful to reduce the inflow of oxygen to cooling system. (3) The bite cutting method for piping in air and press down cutting by roller cutter in the seal bag to prevent inflow of cutting piece, stopper

  7. Cause of defect in the end plug welding of the JOYO fuel pin

    Ouchi, Masaru; Otani, Seiji; Onisi, Koichi; Tateisi, Yoshinori; Ikawa, Yukio.

    1976-01-01

    About twelve thousand fuel pins for the JOYO core fuel were fabricated, and their end plug welding was inspected by X-ray radiography. The defect fractions were 0.2 percent for the lower end plugs and 1.8 percent for the upper, respectively. It had been known that the defect was due to ''line porosity''. In this study, the cause of the ''line porosity defect'' was investigated by the welding experiment performed on some dummy specimens of three different types; open end; closed end; and closed end with dummy pellets and a spring. The position of electrodes was varied for changing the arc gap from 0.3 mm to 1.2 mm. The experimental results are summarized in tables. The results showed that no defect was found in the open end type specimens even with the arc gap of 1.2 mm. Whereas in the other two types of specimens, the defect fraction of 60 to 75 percent was observed with the same arc gap. As for the effect of the arc gap, it was shown that 0.3 mm is the best among 0.3 mm, 0.5 mm and 1.2 mm. No defect was observed in the third type of specimens with the arc gap of 0.3 mm. In summary, it was found that the line porosity defect did not depend on the shape of the end plugs. It is considered to be dependent on both the structure of dummy fuel pins and the position of electrodes. (Aoki, K.)

  8. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    Soga, Tomonori; Ito, Chikara; Aoyama, Takafumi; Suzuki, Soju

    2009-01-01

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x10 15 n/cm 2 s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  9. Flow induced vibration studies for LMFBR in Japan: Past and recent studies of FIV for JOYO and MONJU

    Sato, K [Sodium Engineering Division, O-arai Engineering Centre, Power Reactor and Nuclear Fuel, Development Corporation, Narita-cho, O-arai Machi, Ibaraki-ken (Japan)

    1977-12-01

    This paper presents the past and recent studies of flow induced vibration of the reactor components for the experimental fast breeder reactor JOYO and the prototype fast breeder reactor MONJU, in which many suggestive results for the higher flow velocity systems in a future reactor are contained. The fuel subassembly is the most important from the view point of the vibration. Thus, the studies were carried out with a mock-up subassembly for JOYO. In this experiment, statistical analysis results of the vibration characteristics of single core subassembly and the effects of external forced vibration, flow disturbance and fuel pin bundle vibration were reported. The further more detailed investigations are now being performed for MONJU. In addition to the above studies, the vibration failure of a sodium valve is reported. The valve is a 8-inch stop valve in SODIUM FLOW AND HEAT TRANSFER TEST LOOP at O-arai Engineering Center. The failure occurred in 1969 during the performance test of the mechanical pump, and this resulted in a small sodium leak. The cause of the failure was found to be the vibration fatigue of the metal bellows. (author)

  10. Welding of metallic fuel elements for the irradiation test in JOYO. Preliminary tests and welding execution tests (Joint research)

    Kikuchi, Hironobu; Nakamura, Kinya; Iwai, Takashi; Arai, Yasuo

    2009-10-01

    Irradiation tests of metallic fuels elements in fast test reactor JOYO are planned under the joint research of Japan Atomic Energy Agency (JAEA) and Central Research Institute of Electric Power Industry (CRIEPI). Six U-Pu-Zr fuel elements clad with ferritic martensitic steel are fabricated in Plutonium Fuel Research Facility (PFRF) of JAEA-Oarai for the first time in Japan. In PFRF, the procedures of fabrication of the fuel elements were determined and the test runs of the equipments were carried out before the welding execution tests for the fuel elements. Test samples for confirming the welding condition between the cladding tube and top and bottom endplugs were prepared, and various test runs were carried out before the welding execution tests. As a result, the welding conditions were finalized by passing the welding execution tests. (author)

  11. Effects of activated ACM on expression of signal transducers in cerebral cortical neurons of rats.

    Wang, Xiaojing; Li, Zhengli; Zhu, Changgeng; Li, Zhongyu

    2007-06-01

    To explore the roles of astrocytes in the epileptogenesis, astrocytes and neurons were isolated, purified and cultured in vitro from cerebral cortex of rats. The astrocytes were activated by ciliary neurotrophic factor (CNTF) and astrocytic conditioned medium (ACM) was collected to treat neurons for 4, 8 and 12 h. By using Western blot, the expression of calmodulin dependent protein kinase II (CaMK II), inducible nitric oxide synthase (iNOS) and adenylate cyclase (AC) was detected in neurons. The results showed that the expression of CaMK II, iNOS and AC was increased significantly in the neurons treated with ACM from 4 h to 12 h (PACM and such signal pathways as NOS-NO-cGMP, Ca2+/CaM-CaMK II and AC-cAMP-PKA might take part in the signal transduction of epileptogenesis.

  12. A review of fast reactor program in Japan - April 1984

    Matsuno, Y.

    1984-01-01

    The fast breeder reactor development project in PNC has been in progress steadily in these eighteen years. Concerning the experimental fast reactor, JOYO, the MK-II core attained criticality on November 22, 1982 with 51 fuel assemblies, and received the ''Certificate of Inspection before Operation'' from Government Authority on March 31, 1983, after 100 hours operation with the rated output of 100 MW. Since then, the core has been utilized to implement irradiation bed characteristics test, and to irradiate fuels and structural materials especially for the prototype reactor MONJU. With respect to the prototype reactor MONJU, the installation permit was issued on May 27, 1983, from the prime minister, and the contracts of the first stage between PNC and fabricators were made recently. At the same time, almost all the licenses of preparatory construction works were issued by March 1983, and preparatory construction works were started in April 1983. On the other hand, conceptual design of a demonstration reactor is now under way in a close cooperation with concerned authorities and utilities, as well as investigations of the way of conducting necessary research and development

  13. A review of fast reactor program in Japan (April 2001 - March 2002)

    Nagata, T.; Ieda, Y.

    2002-01-01

    This report describes the research and development activities on fast reactors in Japan thru April 2001 to March 2002. In December 2001, the Cabinet decided the Plan for Reorganization of Government-funded Corporations including the merger of JNC and the Japan Atomic Energy Research Institute (JAERI). A law to set up a new entity is supposed to be submitted to the National Diet by the Japanese Fiscal Year (JFY) 2004. In the Experimental Fast Reactor Joyo, thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. Though a fire broke out in the maintenance building of Joyo in October 2001, the Mk-III construction work was restarted in February 2002. In the Prototype Fast Breeder Reactor Monju, countermeasures against sodium leakage have already been drawn up based on Monju comprehensive safety review. The safety licensing examination for the plant modification of Monju is undergoing. As for the Feasibility Study on Commercialized Fast Reactor Cycle Systems, JFY2001 was the first year of its second phase. A three-year period from JFY2001 to 2003 is the initial term of this phase. During this term, research activities are being focused on the design of the candidate concepts and fundamental tests of key technologies. An interim summary of these activities will be checked and reviewed, and based on the results; the research for JFY 2004 to 2005 will be conducted in order to narrow down the number of alternatives for the fast reactor cycle. (author)

  14. Development of a standard database for FBR core nuclear design (XI). Analysis of the Experimental Fast Reactor 'JOYO' MK-I start-up test and operation data

    Yokoyama, Kenji; Numata, Kazuyuki

    2000-03-01

    As a recent research, Japan Nuclear Cycle Development Institute (JNC) develops a database of 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 the Experimental Fast Reactor 'JOYO' MK-I core. The minimal criticality, sodium void reactivity worth, fuel assembly worth and burn-up coefficient were analyzed. The results of both the minimal criticality and the fuel assembly worth, which were calculated by the standard analytical method for JUPITER experiments, agreed well with the measured values. On the other hand, the results of the sodium void reactivity worth have a tendency to overestimate. As for the burn-up coefficient, it was seen that the C/E values had a dispersion among the operation cycles. The authors judged that further investigation for the estimation of the experimental error will increase the applicability of the integral data to the adjusted library. Furthermore, sensitivity analyses for the minimal criticality, sodium void reactivity worth and fuel assembly worth showed the characteristics of 'JOYO' MK-I core in comparison with ZPPR-9 core of JUPITER experiments. (J.P.N.)

  15. Current development in data acquision and processing system for reactor noise analysis in PUSPATI

    Mohamad Amin Sharifuldin Salleh.

    1986-11-01

    A data acquisition and processing system for reactor noise analysis is described. It consists of four-channel isolation amplifier, a seven-channel DC amplifier, a four-channel analog to digital converter, analog filters, a microcomputer system and a plotter. This system is being applied to investigate the reactor dynamics of the PUSPATI TRIGA MK II reactor. (author)

  16. The Next Step in Somalia: Exploiting Victory Post-Mogadishu

    2012-01-23

    United States. The force was also provided a complete array of equipment to include weapons, body armor and the R21 MkII “Casper” armoured vehicle.47...interdict Al Shabab finances would be to reduce international donations to their fundraising cells. These cells are hidden within the Somali expatriate

  17. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

    Ushiki, Hiroshi; Ito, Hiromichi; Okuda, Eiji; Suzuki, Nobuhiro; Sasaki, Jun; Oota, Katsu; Kawahara, Hirotaka; Takamatsu, Misao; Nagai, Akinori; Okawa, Toshikatsu

    2015-01-01

    In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

  18. IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    experimental series that were performed at 17 different reactor facilities. The Handbook is organized in a manner that allows easy inclusion of additional evaluations, as they become available. Additional evaluations are in progress and will be added to the handbook periodically. Content: FUND - Fundamental; GCR - Gas Cooled (Thermal) Reactor; HWR - Heavy Water Moderated Reactor; LMFR - Liquid Metal Fast Reactor; LWR - Light Water Moderated Reactor; PWR - Pressurized Water Reactor; VVER - VVER Reactor; Evaluations published as drafts 2 - Related Information: International Criticality Safety Benchmark Evaluation Project (ICSBEP); IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments; IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan ; IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database ; IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility; IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation ; IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility ; IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation ; IRPHE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents; IRPHE-ARCH-01, Archive of HTR Primary Documents ; IRPHE/AVR, AVR High Temperature Reactor Experience, Archival Documentation ; IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters; IRPhE/BERENICE, effective delayed neutron fraction measurements ; IRPhE-TAPIRO-ARCHIVE, fast neutron source reactor primary documents, reactor physics experiments. The International Handbook of Evaluated Reactor Physics Benchmark Experiments was prepared by a working party comprised of experienced reactor physics personnel from Belgium, Brazil, Canada, P.R. of China, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Switzerland, United Kingdom, and the United States of America. The IRPhEP Handbook is available to authorised requesters from the

  19. The development of fuel pins and material specimens mixed loading irradiation test rig in the experimental fast reactor Joyo. The development of the fuel-material hybrid rig

    Oyamatsu, Yasuko; Someya, Hiroyuki

    2013-02-01

    In the experimental fast reactor Joyo, there were many tests using the irradiation rigs that it was possible to be set irradiation conditions for each compartment independently. In case of no alternative fuel element to irradiate after unloading the irradiated compartments, the irradiation test was restarted with the dummy compartment which the fuel elements was not mounted. If the material specimens are mounted in this space, it is possible to use the irradiation space effectively. For these reasons, the irradiation rig (hybrid rig) is developed that is consolidated with material specimens compartment and fuel elements compartment. Fuel elements and material specimens differ greatly with heat generation, so that the most important issue in developing of hybrid rig is being able to distribute appropriately the coolant flow which satisfies irradiation conditions. The following is described by this report. (1) It was confirmed that the flow distribution of loading the same irradiation rig with the compartment from which a flow demand differs could be satisfied. (2) It was confirmed that temperature setting range of hybrid rig could be equivalent to that of irradiation condition. (3) By standardizing the coolant entrance structure of the compartment lower part, the prospect which can perform easily recombination of the compartment from which a type differs between irradiation rigs was acquired. (author)

  20. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'

    Ito, Hideaki; Ashida, Takashi; Takamatsu, Misao

    2013-01-01

    Regarding the recovery of fuel exchange capability of 'Joyo', the replacement of Upper Core Structure (UCS) and the retrieval of the sample part of Material Testing Rig With Temperature Control (MARICO-2) are being planned for the fiscal years 2015 and 2016. In the recovery operation, sample part was planned to be removed through the hole created by removing USC because the size of existing rotation plug through-hole is smaller than the size of curved sample part. The procedure outline for this recovery operation was: (1) jack-up UCS, (2) pull out UCS and store in a cask, (3) retrieve sample part, and (4) install a new UCS. In this report, the status of UCS replacement, retrieval of sample part, as well as search and retrieval of loose parts are described. Regarding the replacement of UCS, the topics covered are: (1) removal of adhered sodium, (2) interference of UCS and small rotation plug, (3) the weight of cask for UCS storage, as well as (4) UCS jack-up jig and its functional test. Regarding the retrieval of sample part, the topics are: (1) gripping method selection, (2) pull-up method selection, and (3) safety measures and emergency correspondence. (S.K.)

  1. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. Recovery of MARICO-2 sample part

    Ashida, Takashi; Ito, Hideaki

    2015-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. The following items are introduced here: (1) summary of restoration work and overall process of restoration work, (2) recovery operation of MARICO-2 sample part, (3) exchange of the upper core structure that was conducted this year, and (4) results of recovery of MARIKO-2 sample part. (A.O.)

  2. Evaluation of linear heat rates for the power-to-melt tests on 'JOYO' using the Monte-Carlo code 'MVP'

    Yokoyama, Kenji; Ishikawa, Makoto

    2000-04-01

    The linear heat rates of the power-to-melt (PTM) tests, performed with B5D-1 and B5D-2 subassemblies on the Experimental Fast Reactor 'JOYO', are evaluated with the continuous energy Monte-Carlo code, MVP. We can apply a whole core model to MVP, but it takes very long time for the calculation. Therefore, judging from the structure of B5D subassembly, we used the MVP code to calculate the radial distribution of linear heat rate and used the deterministic method to calculate the axial distribution. We also derived the formulas for this method. Furthermore, we evaluated the error of the linear heat rate, by evaluating the experimental error of the reactor power, the statistical error of Monte-Carlo method, the calculational model error of the deterministic method and so on. On the other hand, we also evaluated the burnup rate of the B5D assembly and compared with the measured value in the post-irradiation test. The main results are following: B5D-1 (B5101, F613632, core center). Linear heat rate: 600 W/cm±2.2%. Burnup rate: 0.977. B5D-2 (B5214, G80124, core center). Linear heat rate: 641 W/cm±2.2%. Burnup rate: 0.886. (author)

  3. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 1. MARICO-2 subassembly retrieval work

    Naito, Hiroyuki; Ashida, Takashi; Ito, Hideaki

    2014-01-01

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. This paper introduces the progress of restoration work and the future work plan, with a focus on the outline of overall restoration work, the method / problems / measures for MARICO-2 sample part recovery operations, and fabrication of sample part recovery device. (A.O.)

  4. The effect of microstructural change on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS) irradiated in JOYO/MARICO-1

    Yano, Yasuhide; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Abe, Yasuhiro

    2004-03-01

    It is well known that the irradiation embrittlement is one of the most important issues to apply ferritic steels for FBR core materials, although ferritic steels have been considered to be candidate core materials of the commercialized FBR core material because of their superior swelling resistance. In order to evaluate the effects of microstructural changes during irradiation on the Charpy impact properties of the high-strength ferritic/martensitic steel (PNC-FMS), microstructural observations were performed with transmission electron microscopy on ruptured halves of the half-sized Charpy specimens of PNC-FMS irradiated in the JOYO/MARICO-1. The results obtained in this study are as follows: (1) There was remarkable disappearance of the lath of martensite in the samples irradiated at 650degC, although there was no significant change in microstructures, especially the lath of martensite between the samples irradiated at 500degC and unirradiated. The disappearance of martensitic lath in the samples irradiated at 650degC was larger than that of the samples thermally aged at 650degC. (2) The ductile-brittle transition temperature (DBTT) of irradiated PNC-FMS is judged to increase with the disappearance of martensitic lath and to decrease with the recovery in dislocations. (3) The decrease in the upper shelf energy (USE) of irradiated PNC-FMS is significantly accompanied by the change of precipitation behavior. (4) The Charpy impact properties and microstructures of PNC-FMS irradiated at 500degC were superior under these irradiation conditions. In future, it is necessary to establish how to evaluate Charpy impact properties in a high fluence region, based on theoretical methods introduced from the data gained in low fluence experiments, in addition to expanding the data area widely. (author)

  5. Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

    Guthrie, J.A.S.

    1980-04-01

    In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

  6. Comparative studies between soja Tshibashi and its two radio-induced mutants

    Kembola Kejuni; Botula Manyala; Kobakozete Itono.

    1978-01-01

    A variety of soybean (Tshibashi 6) has been irradiated by thermal neutrons on a Triga Mk II reactor (flux : 5,5.10 11 n cm -2 :sec.). After irradiation the seeds have been set in an experimental garden, according to the received dose. Comparative studies have been undertaken on the M9 generation. No external significant difference between the mutants and their parents have been noticed

  7. Research work with TRIGA Mark II at the Nuclear Chemistry Section of the 'J. Stefan' Institute in Ljubljana

    Byrne, A.R.; Dermelj, M.; Kosta, L.; Ravkin, V.; Stegnar, P.

    1978-01-01

    The general features of our research programme using TRIGA MK II, as outlined at the last TRIGA Reactor Users Conference in Vienna, September 28-30,1976, remain the same; namely, neutron activation analysis for trace and some minor elements. The four main areas presently investigated are a) environmental studies, b) life sciences research, c) standardization and d) methodology for specific problems arising in the first three topics

  8. Imaging spectroscopy of the terrestrial environment; Proceedings of the Meeting, Orlando, FL, Apr. 16, 17, 1990

    Vane, Gregg

    1990-09-01

    Topics presented include the evolution of the Airborne Visible/Infrared Imaging Spectrometer flight and ground data processing system, direct mineral identification with Geoscan Mk II Advanced Multispectral Scanner, a new approach to imaging spectroscopy, and a linear-wedge spectrometer. Also presented are the effects of moisture content and chemical composition on the near-infrared spectra of forest foliage, change detection in vegetation using 1989 AVIRIS data, and the inversion of high-spectral-resolution data.

  9. PUSPATI TRIGA Reactor

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  10. Berkeley Nuclear Laboratories Reactor Physics Mk. III Experimental Programme. Description of facility and programme for 1971

    Nunn, R M; Waterson, R H; Young, J D

    1971-01-15

    Reactor physics experiments have been carried out at Berkeley Nuclear Laboratories during the past few years in support of the Civil Advanced Gas-Cooled Reactors (Mk. II) the Generating Board is building. These experiments are part of an overall programme whose objective is to assess the accuracy of the calculational methods used in the design and operation of these reactors. This report provides a description of the facility for the Mk. III experimental programme and the planned programme for 1971.

  11. Enterovirus 71 VP1 activates calmodulin-dependent protein kinase II and results in the rearrangement of vimentin in human astrocyte cells.

    Cong Haolong

    Full Text Available Enterovirus 71 (EV71 is one of the main causative agents of foot, hand and mouth disease. Its infection usually causes severe central nervous system diseases and complications in infected infants and young children. In the present study, we demonstrated that EV71 infection caused the rearrangement of vimentin in human astrocytoma cells. The rearranged vimentin, together with various EV71 components, formed aggresomes-like structures in the perinuclear region. Electron microscopy and viral RNA labeling indicated that the aggresomes were virus replication sites since most of the EV71 particles and the newly synthesized viral RNA were concentrated here. Further analysis revealed that the vimentin in the virus factories was serine-82 phosphorylated. More importantly, EV71 VP1 protein is responsible for the activation of calmodulin-dependent protein kinase II (CaMK-II which phosphorylated the N-terminal domain of vimentin on serine 82. Phosphorylation of vimentin and the formation of aggresomes were required for the replication of EV71 since the latter was decreased markedly after phosphorylation was blocked by KN93, a CaMK-II inhibitor. Thus, as one of the consequences of CaMK-II activation, vimentin phosphorylation and rearrangement may support virus replication by playing a structural role for the formation of the replication factories. Collectively, this study identified the replication centers of EV71 in human astrocyte cells. This may help us understand the replication mechanism and pathogenesis of EV71 in human.

  12. Plan and reports of coupled irradiation (JRR-3 and JOYO of research reactors) and hot facilities work (WASTEF, JMTR-HL, MMF and FMF). R and D project on irradiation damage management technology for structural materials of long-life nuclear plant

    Matsui, Yoshinori; Yamamoto, Masaya; Yoshitake, Tsunemitsu; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ichikawa, Shoichi; Yamagata, Ichiro; Soga, Tomonori; Yonekawa, Minoru; Kitamura, Ryoichi; Miyake, Osamu; Takahashi, Hiroyuki; Ishikawa, Kazuyoshi; Kikuchi, Taiji; Usami, Koji; Endo, Shinya; Ichise, Kenichi; Numata, Masami; Onozawa, Atsushi; Aizawa, Masao; Kusunoki, Tsuyoshi; Nakata, Masahito; Abe, Kazuyuki; Ito, Kazuhiro; Takaya, Shigeru; Nagae, Yuji; Wakai, Eiichi; Aoto, Kazumi

    2010-03-01

    'R and D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant' was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of 'Evaluation of Irradiation Damage Indicator' in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research and Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency. (author)

  13. Current status of restoration work for obstacle and supper core structure in reactor vessel of experimental fast reactor 'JOYO'. 3. Sodium purification operation after MARICO recovery and UCS exchange work

    Shimizu, Shunji; Izawa, Osamu; Ishizaki, Kazuhiko; Takeishi, Tsuyoshi; Oowada, Ryohei; Yoshihara, Shizuya; Michino, Masanobu

    2015-01-01

    At fast-breeder reactor 'Joyo', in order to restore the partial inhibition of the rotating plug fuel exchange function due to interference with 'experimental apparatus with instrumentation lines (MARICO-2)', which occurred in May 2007, a recovery work was performed. The replacement work of the upper core structure and the recovery of sample part of the experimental apparatus with instrumentation lines were carried out under conditions where the primary system sodium was drained and the liquid level of reactor vessel was lowered. During the pulling-up work of upper core structure, an increase in nitrogen and hydrogen concentrations in the reactor vessel cover gas (argon) was confirmed through the measurement of the primary system gas chromatograph. This was due to the intrusion of air caused by the opening of the cover gas boundary. Since entrained oxygen reacted with sodium in the reactor, the purity of sodium was reduced. When this sodium is purified according to common method, the sodium with decreased purity defuses through the entire primary cooling system, causing various adverse effects. A safe and reliable procedure to purify sodium while preventing the adverse effects was examined and practiced. (A.O.)

  14. Power-to-melt evaluation of fresh mixed-oxide fast reactor fuel. Technical improvements of the post-irradiation-experiment and the evaluation of the results for the power-to-melt test PTM-2 in 'JOYO'

    Yamamoto, Kazuya; Kushida, Naoya; Koizumi, Atsuhiro

    1999-11-01

    The second Power-To-Melt (PTM) test, PTM-2, was performed in the experimental fast reactor 'JOYO'. All of the twenty-four fuel pins of the irradiation vehicle, B5D-2, for the PTM-2 test, were provided for post-irradiation-experiment (PIE) to evaluate the PTM values. In this study, the PIE technique for PTM test was established and the PTM results were evaluated. The findings are as follows: The maximum fuel-melting ratio on the transverse section was 10.7%, and was within the limit of fuel-melting in this PTM test enough. Unexpected fuel-melting amount to a ratio of 11.8% was found at ∼24 mm below the peak power elevation in a test fuel pin. It is possible that this arose from secondary fuel-melting. Combination of metallographical observation with X-ray microanalysis of plutonium distribution was very effective for the identification of once-molten fuel zone. The PTM evaluation suggested that dependence of the PTM on the fuel pellet density was stronger than that of previous foreign PTM tests, while the dependence on the pellet-cladding gap and the oxygen-to-metal ratio was indistinctly. The dependence on the cladding temperature and the fill gas composition was not shown as well. (author)

  15. Characterizing Mechanisms of Resistance to Androgen Deprivation in Prostate Cancer

    2015-11-01

    of my training in cancer biology has included the attendance of regular seminars and conferences. Indeed, each Tuesday I have attended the meeting of...the Cancer Program of the Broad Institute, and on Tuesday afternoon the institute-wide series of Seminars in Oncology at DFCI, where invited speakers...24(6): p. 719-36. 23. Morris , T.A., R.J. DeLorenzo, and R.M. Tombes, CaMK-II inhibition reduces cyclin D1 levels and enhances the association of

  16. VizieR Online Data Catalog: 315 glitches in the rotation of 102 pulsars (Espinoza+, 2011)

    Espinoza, C. M.; Lyne, A. G.; Stappers, B. W.; Kramer, M.

    2012-02-01

    The Jodrell Bank timing data base comprises observations of more than 700 pulsars, carried out at Jodrell Bank Observatory (JBO) since 1978. Observation procedures are described by Hobbs et al. (2004, Cat. J/MNRAS/353/1311). In summary, these have mostly been performed with the 76-m Lovell telescope, with some complementary observations made using the 30-m MkII and 42-ft telescopes. Every pulsar is observed at typical intervals of 2-10 d in a 64-MHz band centred on 1404-MHz, using an analogue filter bank. Occasionally, observations were also carried out in a band centred at 610-MHz. (3 data files).

  17. La stazione sismica di Serra La Nave sull' Etna

    RIUSCETTI M.

    1967-06-01

    Full Text Available This paper deals with the installation of the first seismological
    station on Mount Etna (Sicily. Three Willmore MK II short
    period seismometers with optical registration have been put into a little
    building, near Astrophysic Observatory at Serra La Nave, on t h e Southern
    side of the mountain. The coordinates of t h e station are: lat. 37°41'30" N,
    long. 14°45'22",9 E, height 1725 m.

  18. Factors controlling physico-chemical characteristics in the coastal waters off Mangalore - A multivariate approach

    Shirodkar, P.V.; Mesquita, A.; Pradhan, U.K.; Verlekar, X.N.; Babu, M.T.; Vethamony, P.

    in the south; Fig.1) using the RCM – 9 MK II current meters, manufactured by Aanderaa Co., Norway. The two locations (13 m and 15 m water depths respectively) were separated by a distance of approximately 10km. Current measurements were carried out for one... and salinity were obtained from different locations using a portable SBE 19 SEACAT Profiler, manufactured by Sea-Bird Electronic, Inc., USA. Vertical profiles were continued for the period 14 -19 April 2007. 2.4. Multivariate statistical analysis a...

  19. Factors affecting the minimum capital cost of a tokamak reactor

    Hancox, R.

    1981-01-01

    The Mk IIA Culham conceptual tokamak reactor design is a 2500 MWe steady-state reactor developed on the basis of a cost optimisation. A revised 1200 MWe conceptual design, the Mk IIB, used a lower wall loading and lower thermodynamic efficiency. A detailed costing of the Mk IIB design, however, showed it to have an unacceptably high capital cost. Since this high cost is a common characteristic of many fusion reactor designs, the cost optimisation of the Mk II design has been reconsidered. (author)

  20. Radiological monitoring related to the operation of PUSPATI's Triga Reactor

    Fatimah Mohamad Amin; Mohamad Yusof Mohamad Ali; Lau How Mooi; Idris Besar.

    1983-01-01

    Reactor operation is one of the main activities carried out at the Tun Ismail Atomic Research Centre (PUSPATI) which requires radiological monitoring. This paper describes the programme for radiological monitoring which is related to the operation of the 1 MW Triga MK II research reactor which was commissioned in July, 1982. This programme includes monitoring of the radiation and contamination levels of the reactor and its associated facilities and environmental monitoring of PUSPATI's site and its environs. The data presented in this paper covers the period between 1982 to 1983 which includes both the pre-operational and operational phases of the monitoring programme. (author)

  1. Numerical studies of large penetrations and closures for containment vessels subjected to loadings beyond the design basis

    Kulak, R.F.; Hsieh, B.J.; Kennedy, J.M.; Ash, J.E.; McLennan, G.A.

    1984-01-01

    Numerical simulations of the macro-deformations of the sealing surfaces (gasketed junctures) of a PWR steel containment vessel's equipment hatch and a BWR Mk II containment vessel head have been performed. Results for the equipment hatch juncture indicate that the rotations of the hatch cover and penetration sleeve must be accounted for when performing leakage analysis because they can effect the compression of the gasket even though the gasket is in a pressure-seated configuration. Results from a leakage analysis indicated that excessive leakage can occur if the surface roughness is high and/or the compression set is high. Results for the Mk II head show that both the temperature and pressure loadings must be taken into account to obtain realistic responses. The temperature difference between the flanges and bolts has the important net effect of keeping the gasketed juncture closed, that is in metal-to-metal contact. Due to the high accident temperature, the gasket itself was found to achieve 100% compression set and thus could not perform its sealing function within the juncture

  2. Description of steam-condensation phenomena during the loss-of-coolant accident

    McCauley, E.W.; Holman, G.S.; Aust, E.; Schwan, H.; Vollbrandt, J.; Fuerst, H.

    1980-01-01

    The development and verification of advanced computer models which describe the boiling water reactor (BWR) pressure suppression process for a hypothetical loss-of-coolant accident (LOCA) require a clear description of basic steam condensation phenomena. The GKSS Research Center, in coordination with interested institutions of West Germany and the United States, is currently conducting a test program for such basic research on a multivent BWR-related pressure suppression system. The Lawrence Livermore National Laboratory (LLNL) acts as the principal US NRC liaison for this test program, with particular emphasis on development of GKSS data for confirmatory use regarding US Mark II nuclear power plants as well as to advanced code development. The multivent test facility, placed in operation in February 1979, is a three-pipe full-scale vent system modelling main features of both the West German KWU and United States G.E. Mk II BWR pressure suppression systems. The test facility and testing programs are described

  3. Narrow power deposition profiles on the JET divertor target

    Lingertat, J.; Laux, M.; Monk, R.

    2001-01-01

    One of the key unresolved issues in the design of a future fusion reactor is the power handling capability of the divertor target plates. Earlier we reported on the existence of narrow power deposition profiles in JET, obtained mainly from Langmuir probe measurements. We repeated these measurements in the MkI, MkII and MkIIGB divertor configurations with an upgraded probe system, which allowed us to study the profile shape in more detail. The main results of this study are: In NB heated discharges the electron temperature and power flux at the outer target show a distinct peak of ∼5 mm half-width near the separatrix strike point. The corresponding profiles on the inner target do not show a similar feature. The height of the narrow peak increases with NB heating power and decreases with deuterium and impurity gas puffing. Ion orbit losses are suggested as a possible explanation of the observed profile shape

  4. Station blackout calculations for Peach Bottom

    Hodge, S.A.

    1985-01-01

    A calculational procedure for the Station Blackout Severe Accident Sequence at Browns Ferry Unit One has been repeated with plant-specific application to one of the Peach Bottom Units. The only changes required in code input are with regard to the primary continment concrete, the existence of sprays in the secondary containment, and the size of the refueling bay. Combustible gas mole fractions in the secondary containment of each plant during the accident sequence are determined. It is demonstrated why the current state-of-the-art corium/concrete interaction code is inadequate for application to the study of Severe Accident Sequences in plants with the BWR MK I or MK II containment design

  5. Culham Conceptual Tokamak Mark II. Design study of the layout of a twin-reactor fusion power station

    Guthrie, J.A.S.; Harding, N.H.

    1981-07-01

    This report describes the building layout and outline design for the nuclear complex of a fusion reactor power station incorporating two Culham Conceptual Tokamak Reactors Mk.II. The design incorporates equipment for steam generation, process services for the fusion reactors and all facilities for routine and non-routine servicing of the nuclear complex. The design includes provision of temporary facilities for on site construction of the major reactor components and shows that these facilities may be used for disassembly of the reactors either for major repair and/or decommissioning. Preliminary estimates are included, which indicate the cost benefits to be obtained from incorporating two reactors in one nuclear complex and from increased wall loading. (author)

  6. A survey of formaldehyde in the Cepheus OB3 molecular cloud

    Few, R.W.; Cohen, R.J.

    1983-01-01

    The 1 11 - 1 10 absorption line of formaldehyde at 6-cm wavelength has been surveyed over the region of the Cepheus OB3 molecular cloud, using the Jodrell Bank Mk II radio telescope (beamwidth 9 x 10 arcmin 2 ). The measurements have a velocity resolution of 0.27 km s - 1 and an rms noise level of approx. 0.01 K. The formaldehyde has a very clumpy distribution which is broadly similar to the CO distribution found by Sargent. A total molecular mass of 1.9 x 10 4 solar masses is implied by the formaldehyde measurements. Cepheus A is not the dominant concentration in the formaldehyde map. The most massive formaldehyde concentration is Cepheus C, which has a mass of 3600 solar masses. It appears to be stabilized by rotation. (author)

  7. The nuclear fuel cycle: (2) fuel element manufacture

    Doran, J.

    1976-01-01

    Large-scale production of nuclear fuel in the United Kingdom is carried out at Springfields Works of British Nuclear Fuels Ltd., a company formed from the United Kingdom Atomic Energy Authority in 1971. The paper describes in some detail the Springfields Works processes for the conversion of uranium ore concentrate to uranium tetrafluoride, then conversion of the tetrafluoride to either uranium metal for cladding in Magnox to form fuel for the British Mk I gas-cooled reactors, or to uranium hexafluoride for enrichment of the fissile 235 U isotope content at the Capenhurst Works of BNFL. Details are given of the reconversion at Springfields Works of this enriched uranium hexafluoride to uranium dioxide, which is pelleted and then clad in either stainless steel or zircaloy containers to form the fuel assemblies for the British Mk II AGR or advanced gas-cooled reactors or for the water reactor fuels. (author)

  8. The feasibility of remotely separating and rejoining the main coolant pipes of a fusion reactor

    Briaris, D.A.; Stanbridge, J.R.

    1977-09-01

    The generic requirement of a fusion reactor that the first wall and other high neutron dose structures be periodically replaced gives rise to a number of complex engineering operations which need to be performed remotely and with a high degree of reliability. Techniques for the remote separation and rejoining of the helium coolant pipes on the Culham Conceptual Tokamak Reactor Mk. II have been investigated in the form of cutting and welding schemes and the use of a mechanical coupling. A mechanical coupling is the more attractive because the reduced complexity of the operations to separate and join the pipes potentially shortens the reactor down-time. Some assessment of remote joint examination and recovery from faults has also been made. (author)

  9. Feature test report for the Small Debris Collection and Packaging System

    Brisbin, S.A.

    1995-01-01

    The Spent Nuclear Fuel Equipment Engineering group performed feature testing of the Small Debris Collection and Packaging System (SDCPS) in the 305 Cold Test Facility from January 30, 1995, to February 1, 1995. Feature testing of the Small Debris Collection and Packaging System (SDCPS) was performed for the following reasons: To assess the feasibility of using ''drop-out'' vessels to collect small debris (<2.5 cm) in MK-II fuel canisters while transferring sludge to the Weasel Pit. To evaluate system performance under conditions similar to those in the K-Basins (e.g. submerged under 4.9 meters of water and operated with long handled tools) while using a surrogate sludge mixed with debris. To determine if canister weight could be used to predict the volume of sludge and/or debris contained within the canisters during system operation

  10. Observations of the Magellanic Stream between declinations -200 and 00

    Cohen, R.J.

    1982-01-01

    The region of the Magellanic Stream between RA 23sup(h) 00sup(m) and 00sup(h) 20sup(m) and Dec - 20 0 and 0 0 (1950) has been mapped in the 21-cm line of neutral hydrogen using the Jodrell Bank Mk II telescope (beamwidth 31 x 34 arcmin 2 ). The detection level of the measurements is 0.1 K. The Stream is much more extensive in this part of the sky than hitherto realized, and has a very complex filamentary structure. All the filaments follow a regular velocity pattern. In addition to the known gradient of velocity along the Stream there is a gradient transverse to the Stream. In this and other respects the Stream is very similar to tidal bridges and tails seen in the nearby M81 group of galaxies. (author)

  11. Malaysian perspective on the contribution of nuclear science and technology to national development

    Alang Md Rashid, Nahrul Khair [Unit Tenaga Nuklear, Bangi, Selangor (Malaysia)

    1994-04-01

    The development of nuclear science and technology in Malaysia began with the inception of The Nuclear Energy Unit (UTN) in 1972. In 1985, the Atomic Energy Licensing Board was set up as a regulatory body to enforce the Atomic Energy Licensing Act. Ten years after UTN`s establishment, the first of its major facilities, a one Megawatt TRIGA MkII nuclear research reactor (RTP), was commissioned. This is the first step of any type of nuclear reactor for Malaysia. The healthy development of peaceful uses of nuclear science and technology in malaysia has enabled UTN to acquire several other major facilities. These facilities support research and development, in line with UTN`s mission, viz, to enhance national development through the applications of nuclear science and technology. This paper describes selected activities at UTN and some of its successes in linking the results of research and development to real-world applications through services and/or technology transfers.

  12. Malaysian perspective on the contribution of nuclear science and technology to national development

    Nahrul Khair Alang Md Rashid

    1994-01-01

    The development of nuclear science and technology in Malaysia began with the inception of The Nuclear Energy Unit (UTN) in 1972. In 1985, the Atomic Energy Licensing Board was set up as a regulatory body to enforce the Atomic Energy Licensing Act. Ten years after UTN's establishment, the first of its major facilities, a one Megawatt TRIGA MkII nuclear research reactor (RTP), was commissioned. This is the first step of any type of nuclear reactor for Malaysia. The healthy development of peaceful uses of nuclear science and technology in malaysia has enabled UTN to acquire several other major facilities. These facilities support research and development, in line with UTN's mission, viz, to enhance national development through the applications of nuclear science and technology. This paper describes selected activities at UTN and some of its successes in linking the results of research and development to real-world applications through services and/or technology transfers

  13. Rare earth elements determination and distribution patterns in sediments of polluted marine environment by instrumental neutron activation analysis

    Akyil, S.; Yusof, A.M.; Wood, A.K.H.

    2001-01-01

    Results obtained from the analysis of sediment core samples taken from a fairly polluted marine environment were analyzed for the REE contents to determine the concentrations of La, Ce, Sm, Eu, Tb, Dy and Yb using instrumental neutron activation analysis. Core samples were divided into strata of between 2 to 3 cm intervals and prepared in the powdered form before irradiating them in a TRIGA Mk.II reactor. Down-core concentration profiles of La, Ce, Sm, Eu, Tb, Dy and Yb in 3 core sediments from three sites are obtained. The shale-normalized REE pattern from each site was examined and later used to explain the history of sedimentation by natural processes such as shoreline erosion and weathering products deposited on the seabed and furnishing some baseline data and/or pollution trend occurring within the study area

  14. Official Guard and Reserve Manpower Strengths and Statistics: FY 1989 Summary

    1989-01-01

    S0- 5.8 co It0. 0 m a* t7 N 𔃾 in o𔃾 t N o - n-K, 0 mI- IZ 0 N m i an N 0o cm a, . KZK UK 14i N~’ 47 𔃾 N 0’ 0 D- 0 00a, c KlS K K I NNN 7n 𔃾 𔃾 No...W C, 00C0F)a*P m 00 on 0..0.....P00Mlso0 do w .P- 638%1 P 1 KZK ~ ~ P)goMK II- .8 4N i 40. 8 4 I.5M!.4M0󈧣U . * . P) woo- 3. 30 K 30A- N C4 in4C i K

  15. The Expression of Can and Camk is Associated with Lipogenesis in the Muscle of Chicken

    Y Yang

    2015-09-01

    Full Text Available ABSTRACTIntramuscular fat (IMF content in chickens significantly contributes to meat quality. The main objective of this study was to assess the expression of calcineurin (CaN and Ca2+/calmodulin-dependent protein kinase (CaMK in lipogenesis in chicken muscle. Chickens were slaughtered and sampled at 4, 8, and 16 weeks of age. IMF content and the expression of CaN subunits and CaMK isoforms were measured in the thigh muscle tissue. The results showed that the IMF contents were greater at 16 weeks compared with those at 4 and 8 weeks (p<0.05. Transcription of fatty acid synthase (FAS and fatty acid translocase CD36 (FAT/CD36 mRNA significantly increased with age, from four to 16 weeks (p<0.05. The mRNA levels of CaNB and CaMK IV were significantly lower at 16 weeks than at four weeks (p<0.05, but CaMK II mRNA levels were significantly higher than at four weeks (p<0.05. In order to evaluate the role of CaMK and CaN in adipogenesis, SV cells were incubated in standard adipogenic medium for 24 h and treated with specific inhibitor of CaMK and CaN. The expressions of CCAAT/enhancer binding protein b (C/EBPb, sterol regulatory element-binding protein 1 (SREBP1,and peroxisome proliferation-activated receptor g (PPARγwere dramatically enhanced by the CsA, CaN inhibitor (p<0.05. KN93, CaMK II inhibitor, dramatically repressed the expression of those lipogenic gene (p<0.05. These results indicated that CaN and CaMK had different effects on adipogenesis in the muscle of chickens.

  16. Characterization of a calcium/calmodulin-dependent protein kinase homolog from maize roots showing light-regulated gravitropism

    Lu, Y. T.; Hidaka, H.; Feldman, L. J.

    1996-01-01

    Roots of many species respond to gravity (gravitropism) and grow downward only if illuminated. This light-regulated root gravitropism is phytochrome-dependent, mediated by calcium, and inhibited by KN-93, a specific inhibitor of calcium/calmodulin-dependent protein kinase II (CaMK II). A cDNA encoding MCK1, a maize homolog of mammalian CaMK, has been isolated from roots of maize (Zea mays L.). The MCK1 gene is expressed in root tips, the site of perception for both light and gravity. Using the [35S]CaM gel-overlay assay we showed that calmodulin-binding activity of the MCK1 is abolished by 50 microM KN-93, but binding is not affected by 5 microM KN-93, paralleling physiological findings that light-regulated root gravitropism is inhibited by 50 microM KN-93, but not by 5 microM KN-93. KN-93 inhibits light-regulated gravitropism by interrupting transduction of the light signal, not light perception, suggesting that MCK1 may play a role in transducing light. This is the first report suggesting a physiological function for a CaMK homolog in light signal transduction.

  17. Effect of Neutron Irradiation on the Physicochemical Properties of Naproxen Sodium

    Ibrahim Ijang

    2016-01-01

    Complex dosage forms may be designed to provide sustained release of the drug or to deliver the active ingredients to the specific sites. It is important to know the in-vivo behaviour of the drug formulation following administration. Gamma scintigraphy technique has been widely used to monitor the in-vivo radiopharmaceuticals dosage form by neutron activation. This study was to investigate effect of neutron activation on the physicochemical properties of the Naproxen Sodium as a model drug. The drug was irradiated using TRIGA MK II reactor with thermal neutron at 1.2 x 10"1"2 neutron cm"-"2s"-"1 for 1, 2, 3, 4, 5 and 30 minutes. The stability of naproxen sodium was assessed based on the malting point, morphology, Gas Chromatograph-Mass Spectrometer (GC-MS) and Fourier Transform Infrared Spectrometer (FTIR). Results of analysis of Scanning Electron Microscope (SEM) and FTIR showed changes in the physicochemical properties of naproxen sodium when duration of irradiation was increased. There were no major changes in the result of GC-MS and Differential Scanning Calorimeter (DSC). Based on the results obtained, it be concluded that naproxen sodium is a suitable drug that can be used for neutron activation based gamma scintigraphy. The maximum irradiation time that naproxen sodium can be withstand without changes in its physicochemical properties is 3 minutes. (author)

  18. Role of BWR secondary containments in severe accident mitigation: issues and insights from recent analyses

    Greene, S.R.

    1988-01-01

    All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by a secondary containment consisting of a reactor building and refueling bay (MK I and MK II designs), a shield building, auxiliary building and fuel building (MK III), or an auxiliary building and enclosure building (Grand Gulf style MK III). Although secondary containment designs are highly plant specific, their purpose is to minimize the ground level release of radioactive material for a spectrum of traditional design basis accidents. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident. This paper presents a brief overview of domestic BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent ORNL secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented

  19. Formaldehyde in the Galactic Centre

    Cohen, R.J.; Few, R.W.

    1981-01-01

    Formaldehyde 6-cm absorption in the direction of the Galactic Centre has been surveyed using the Jodrell Bank MK II radio telescope (beam-width 10 x 9 arcmin). The observations sample the region - 2 0 = 0 and - 0 0 .5 = 0 .5, with a velocity range of 620 km s -1 , a velocity resolution of 2.1 km s -1 and an rms noise level of approximately 0.03 K. The data are presented as contour maps showing line temperature as a function of latitude and velocity (b-V maps) and as a function of longitude and velocity (l-V maps). Similar maps of the line-to-continuum ratio are also presented. The radial distribution of formaldehyde (H 2 CO) in the Galactic Centre region is derived using two different kinematic models which give similar results. Formaldehyde is strongly concentrated in the Galactic Centre in a layer of latitude extent approximately 0 0 .5 and longitude extent approximately 4 0 which contains one quarter of all the H 2 CO in the Galaxy. The distribution is centred on l approximately 1 0 . The individual H 2 CO features are described in detail. (author)

  20. A survey of high-velocity H I in the Cetus region

    Cohen, R.J.

    1982-01-01

    The region 02sup(h) 16sup(m) 0 0 surrounding the Cohen and Davies complex of high-velocity clouds has been surveyed in the 21-cm line of H I using the Jodrell Bank MK II radio telescope (beamwidth 31 x 34 arcmin). The high-velocity cloud complex was sampled every 2sup(m) in right ascension and every 0 0 .5 in declination. The observations cover a velocity range of 2100 km s -1 with a resolution of 7.3 km s -1 and an rms noise level of 0.025 K. No HVCs were found outside the velocity range -400 to +100 km s -1 . The data are presented on microfiche as a set of contour maps showing 21-cm line temperature as a function of declination and radial velocity at constant values of right ascension. Discussion is centred on the very-high-velocity clouds at velocities of -360 to -190 km s -1 . It is concluded that they are probably debris from the tidal interaction between our Galaxy and the Magellanic Clouds. (author)

  1. Installation and sampling of vadose zone monitoring devices

    Bergeron, S.M.; Strickland, D.J.; Pearson, R.

    1987-10-01

    A vadose zone monitoring system was installed in a sanitary landfill near the Y-12 facility on the Department of Energy's Oak Ridge, Tennessee Reservation. The work was completed as part of the LLWDDD program to develop, design, and demonstrate new low level radioactive waste disposal monitoring methods. The objective of the project was to evaluate the performance of three types of vadose zone samplers within a similar hydrogeologic environment for use as early detection monitoring devices. The three different types of samplers included the Soil Moisture Equipment Corporation Pressure-Vacuum samplers (Models 1920 and 1940), and the BAT Piezometer (Model MK II) manufactured by BAT Envitech, Inc. All three samplers are designed to remove soil moisture from the vadose (unsaturated) zone. Five clusters of three holes each were drilled to maximum depths of 45 ft around part of the periphery of the landfill. Three samplers, one of each type, were installed at each cluster location. Water samples were obtained from 13 of the 15 samplers and submitted to Martin Marietta for analysis. All three samplers performed satisfactorily when considering ease of installation, required in-hole development, and ability to collect water samples from the vadose zone. Advantages and disadvantages of each sampler type are discussed in the main report

  2. Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

    Hancox, R.; Mitchell, J.T.D.

    1977-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

  3. Antioxidant hydrolysed peptides from Manchurian walnut (Juglansmandshurica Maxim.) attenuate scopolamine-induced memory impairment in mice.

    Ren, Dayong; Zhao, Fanrui; Liu, Chunlei; Wang, Ji; Guo, Yong; Liu, Jingsheng; Min, Weihong

    2018-04-13

    Walnut protein, which is obtained as a by-product of oil expression, has not been used efficiently. Although walnuts are beneficial for cognitive functioning, the potential of their protein composition in strengthening learning and memory functions remains unknown. In this research, the inhibition of memory impairment by the Manchurian walnut hydrolyzed peptide (MWHP) was evaluated. Small-molecular-weight MWHP (<3 kDa) achieved the optimal antioxidative activity. Therefore, MWHP (<3 kDa) was subjected to the following mice trials to evaluate its attenuation effect on memory impairment. In the Morris water maze test, MWHP shortened the total path for searching the platform, reduced the escape latency, and increased the dwelling distance and time in the coverage zone. MWHP also prolonged the latency and diminished errors in the passive avoidance response tests. These behavioral tests demonstrated that MWHP could inhibit scopolamine-induced memory impairment. MWHP improved memory by reducing oxidative stress, inhibiting apoptosis, regulating neurotransmitter functions, maintaining hippocampal CA3 pyramidal neurons, and increasing p-CaMK II levels in brain tissues. Experimental results proved that MWHP exhibits potential in improving memory and should be used to develop novel functional food. This article is protected by copyright. All rights reserved.

  4. On the development of radiation tolerant surveillance camera from consumer-grade components

    Klemen Ambrožič

    2017-01-01

    Full Text Available In this paper an overview on the process of designing a radiation tolerant surveillance camera from consumer grade components and commercially available particle shielding materials is given. This involves utilization of Monte-Carlo particle transport code MCNP6 and ENDF/B-VII.0 nuclear data libraries, as well as testing the physical electrical systems against γ radiation, utilizing JSI TRIGA mk. II fuel elements as a γ-ray sources. A new, aluminum, 20 cm × 20 cm × 30 cm irradiation facility with electrical power and signal wire guide-tube to the reactor platform, was designed and constructed and used for irradiation of large electronic and optical components assemblies with activated fuel elements. Electronic components to be used in the camera were tested against γ-radiation in an independent manner, to determine their radiation tolerance. Several camera designs were proposed and simulated using MCNP, to determine incident particle and dose attenuation factors. Data obtained from the measurements and MCNP simulations will be used to finalize the design of 3 surveillance camera models, with different radiation tolerances.

  5. On the development of radiation tolerant surveillance camera from consumer-grade components

    Klemen, Ambrožič; Luka, Snoj; Lars, Öhlin; Jan, Gunnarsson; Niklas, Barringer

    2017-09-01

    In this paper an overview on the process of designing a radiation tolerant surveillance camera from consumer grade components and commercially available particle shielding materials is given. This involves utilization of Monte-Carlo particle transport code MCNP6 and ENDF/B-VII.0 nuclear data libraries, as well as testing the physical electrical systems against γ radiation, utilizing JSI TRIGA mk. II fuel elements as a γ-ray sources. A new, aluminum, 20 cm × 20 cm × 30 cm irradiation facility with electrical power and signal wire guide-tube to the reactor platform, was designed and constructed and used for irradiation of large electronic and optical components assemblies with activated fuel elements. Electronic components to be used in the camera were tested against γ-radiation in an independent manner, to determine their radiation tolerance. Several camera designs were proposed and simulated using MCNP, to determine incident particle and dose attenuation factors. Data obtained from the measurements and MCNP simulations will be used to finalize the design of 3 surveillance camera models, with different radiation tolerances.

  6. High performance with modified magnetic shear in JET DD and DT plasmas

    Gormezano, C.

    1999-01-01

    Internal transport barriers (ITBs) in which both the ion thermal diffusivity and electron thermal diffusivity are substantially reduced have been observed in JET. Such discharges have been obtained with DD and DT plasmas. Central ion temperatures of 40 keV and plasma pressure gradients of 10 6 Pa/m were observed in DT plasmas leading to a fusion triple product n i0 T i0 τ E 1.1 x 10 21 m -3 ·keV·s and producing 8.2 MW of fusion power. ITBs have been produced in both the MkII and the new Gas Box divertor configuration with similar behaviour. With the Gas Box divertor an L mode edge has so far only been produced using edge radiation cooling. For the first time, ITBs have been triggered by radiating about 40% of the power with a krypton puff. A possible scaling of the power needed to trigger an ITB with magnetic field is suggested. (author)

  7. High performance with modified shear in JET D-D and D-T plasmas

    2001-01-01

    The observation of Internal Transport Barriers (ITBs) in which ion thermal diffusivity is reduced to a neo- classical level and the electron thermal diffusivity is substantially reduced has been made in JET with the optimised shear scenario with the MkII divertor both in D-D and in D-T. Central ion temperatures of 40keV and plasma pressure gradient of 10 6 Pa/m were observed in D-T leading to a fusion triple product n i T i τ E =1x10 21 m -3 keVs and 8.2MW of fusion power. ITBs have also been produced in the new Gas Box divertor configuration with a similar behaviour. With the new divertor an L-mode edge has only been produced using edge radiation cooling. For the first time, ITBs have been triggered by radiating about 40% of the power with a krypton puff. A tentative scaling of the power needed to trigger an ITB with magnetic field is indicated. (author)

  8. Comparison of three facebow/semi-adjustable articulator systems for planning orthognathic surgery.

    O'Malley, A M; Milosevic, A

    2000-06-01

    Our aim was to measure the steepness of the occlusal plane produced by three different semi-adjustable articulators: the Dentatus Type ARL, Denar MkII, and the Whipmix Quickmount 8800, and to assess the influence of possible systematic errors in positioning of study casts on articulators that are used to plan orthognathic surgery. Twenty patients (10 skeletal class II, and 10 skeletal class III) who were having pre-surgical orthodontics at Liverpool University Dental Hospital were studied. The measurement of the steepness of the occlusal plane was taken as the angle between the facebow bite-fork and the horizontal arm of the articulator. This was compared with the angle of the maxillary occlusal plane to the Frankfort plane as measured on lateral cephalometry (the gold standard). The Whipmix was closest to the gold standard as it flattened the occlusal plane by only 2 degrees (P<0.05). The results of the Denar and Dentatus differed significantly from those of the cephalogram as they flattened the occlusal plane by 5 degrees and 6. 5 degrees (P<0.01), respectively. Clinicians are encouraged to verify the steepness of the occlusal plane on mounted study casts before the technician makes the model. Copyright 2000 The British Association of Oral and Maxillofacial Surgeons.

  9. Pulsed negative hydrogen source for currents up to one ampere

    Prelec, K.; Sluyters, T.

    1975-01-01

    During the 2nd Symposium on Ion Sources and Formation of Ion Beams, the development of a Mk II pulsed double slit magnetron source for the production of negative hydrogen ions was discussed. The source was capable of yielding beam currents up to 125 milliamperes, corresponding to current densities of 1.25 A/cm 2 . In order to increase negative hydrogen beam intensities by an order of magnitude (this would be quite useful for initial high energy neutral injector systems on Tokamaks), a larger, Mk III magnetron has been constructed, with the number of slits increased up to six. The idea was to utilize in a more efficient way the plasma width. In addition, such a source geometry will be more adaptable for beam formation and acceleration than single slit structures. With three extraction slits, a negative hydrogen yield of 300 mA was obtained with current densities of 1.2 A/cm 2 ; preliminary results with six extraction slits showed beam currents in excess of half an ampere with averaged current densities in excess of 0.75 A/cm 2 . (U.S.)

  10. Reactor costs and maintenance, with reference to the Culham Mark II conceptual Tokamak reactor design

    Hancox, R.; Mitchell, J.T.D.

    1976-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, due mainly to the low power density of the fusion reactor which affects both the reactor and building costs. In order to reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (orig.) [de

  11. Determination of the clean 4f peak shape in XPS for plutonium metal

    Morrall, P. [AWE, Aldermaston, Reading, Berkshire RG7 4PR (United Kingdom)], E-mail: peter.morrall@awe.co.uk; Roussel, P. [AWE, Aldermaston, Reading, Berkshire RG7 4PR (United Kingdom); Jolly, L.; Brevet, A.; Delaunay, F. [Commissariat a l' Energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)

    2009-03-15

    Many of the interesting properties observed with plutonium are ascribed to the influence of 5f electrons, and to the degree of localisation observed within these electrons. Indeed, changes in 5f localisation are sensitively reflected in the final states observed in core-level photoemission measurements. However, when analysing the 4f manifold of elemental plutonium, it is essential to obtain spectra without the influence of oxidation, which can easily be misinterpreted as 5f localisation. The ideal method to extract elemental plutonium 4f spectra is to remove any influence of oxidation from the 'clean' plutonium data by careful measurement of the oxygen 1s region, and the subsequent subtraction of the unwanted oxide features. However, in order to achieve this objective it is essential to determine the relative sensitivity factor (RSF) for plutonium 4f and the precise shape of the 4f features from plutonium sesqui-oxide. In this paper, we report an experimental determination of the RSF for the plutonium 4f manifold using experimental data captured from two different Vacuum Generators spectrometers; an ESCALAB Mk II and an ESCALAB 220i.

  12. Cutting-edge analysis of extracellular microparticles using ImageStream(X) imaging flow cytometry.

    Headland, Sarah E; Jones, Hefin R; D'Sa, Adelina S V; Perretti, Mauro; Norling, Lucy V

    2014-06-10

    Interest in extracellular vesicle biology has exploded in the past decade, since these microstructures seem endowed with multiple roles, from blood coagulation to inter-cellular communication in pathophysiology. In order for microparticle research to evolve as a preclinical and clinical tool, accurate quantification of microparticle levels is a fundamental requirement, but their size and the complexity of sample fluids present major technical challenges. Flow cytometry is commonly used, but suffers from low sensitivity and accuracy. Use of Amnis ImageStream(X) Mk II imaging flow cytometer afforded accurate analysis of calibration beads ranging from 1 μm to 20 nm; and microparticles, which could be observed and quantified in whole blood, platelet-rich and platelet-free plasma and in leukocyte supernatants. Another advantage was the minimal sample preparation and volume required. Use of this high throughput analyzer allowed simultaneous phenotypic definition of the parent cells and offspring microparticles along with real time microparticle generation kinetics. With the current paucity of reliable techniques for the analysis of microparticles, we propose that the ImageStream(X) could be used effectively to advance this scientific field.

  13. MK-III function tests in JOYO. Primary main cooling pump

    Isozaki, Kazunori; Saito, Takakazu; Sumino, Kouzo; Karube, Kouji; Terano, Toshihiro; Sakaba, Hideo; Nakai, Satoru

    2004-06-01

    MK-III function test (SKS-1) that was carried out from October 17, 2001 through October 23, 2001 using MK-III transition core configuration and MK-III function tests (SKS-2) was carried out from January 27, 2003 through February 13, 2003 using MK-III core configuration. The major function tests results of primary cooling system were shown as follows; (1) The stability of the primary main pump flow control system was confirmed on both CAS (cascade) mode and Man (manual) mode. Also no divergence of flow and revolution of the pump were observed at step flow change disturbance. (2) The main motor was shifted to run-back flow control operation in about 54 seconds after scram. The flow rate and pump revolution at run-back operation of A and B cooling system were 167 m 3 /h and 117 rpm, 185m 3 /h and 118 rpm respectively. The pump revolution was within the design target revolution 122 rpm ± 8 rpm and the flow was over the 10% of the rated flow. (3) The pony motor was engaged in operation in about 39 seconds after the primary main pump trip. The flow rate and pump revolution at the pony motor operation of A and B cooling system were 180 m 3 /h and 124 rpm, 190 m 3 /h and 123 rpm respectively. These values were satisfied the design low limit of 93 rpm and 10% of the rated flow. (4) Free flow coast down time constant was longer than 10 seconds that was design shortest time at both the primary pump trip and run-back operation. (5) Pump over flow column sodium levels of both A and B cooling system at rated operating condition were NL-1550 mm and, NL-1468 mm respectively and were lower than NL-1581 mm of the design value. This result shows the new IHX pressure loss estimation was conservative. (6) It was confirmed that the primary main pump could operate with out scram for up to 0.6 seconds of external power supply loss. (author)

  14. Periodic safety review of the experimental fast reactor JOYO. Review of aging management

    Isozaki, Kazunori; Ogawa, To-ru; Nishino, Kazunari

    2005-05-01

    Periodic safety review (Review of the aging management) which consisted of ''Technical review on aging for the safety related structures, systems and components'' and ''Establishment a long term maintenance program'' was carried out up to April 2005. 1. Technical review on aging for the safety related structures, systems and components. It was technically confirmed to prevent the loss of function of the safety related structures, systems and components due to aging phenomena, which (1) irradiation damage, (2) corrosion, (3) abrasion and erosion, (4) thermal aging, (5) creep and fatigue, (6) Stress Corrosion Cracking, (7) insulation deterioration and (8) general deterioration, under the periodic monitoring or renewal of them. 2. Establishment of long term maintenance program. The long term maintenance during JFY2005 to 2014 were established based on the technical review on aging for the safety related structures, systems and components. It was evaluated that the inspection and renewal based on the long term maintenance program, in addition to the spontaneous inspection of the long term voluntary long-term inspection plan, could prevent the loss of function of the safety related structures, systems and components. (author)

  15. The off-line computation system for supervising performance of JOYO: JOYPAC system, 2

    Suzuki, Tomoo; Hasegawa, Akira; Akimoto, Masayuki; Miyamoto, Yoshiaki; Katsuragi, Satoru

    1976-10-01

    HONEYCOMB is a code for detailed calculations in analyzing nuclear characteristics of the reactor. It performs criticality calculation in diffusion model and burn up calculation, for 3-dimensional hexagonal-z geometry. It can predict the critical insertion depth of control rods and calculate the 3-dimensional power distribution required by thermo-hydraulic calculation. Power distribution and burn up are also obtained for fuel pins, if necessary, as well as for assemblies. FDCAL-2 predicts coolant flow distribution in every coolant channel between inlet and outlet plenums in the reactor vessel. In calculating the flow distribution in the assemblies, the subchannel model is used, and the thermal mixing effect is expressed in terms of an apparent heat transfer coefficient. FATEC-3 calculates temperature distribution within some assemblies, optionally specified in the given core matrix. At the same time, it estimates the hot-spot temperature, one of the informations for confirming the safe operation. FACAL-2 and FATEC-3 have been combined so as to remove their unnecessary overlapping parts, and have consequently formed a detailed calculation code for analyzing thermo-hydraulic characteristics of the reactor, FDCAL-3. FDCAL-3 has been linked to HONEYCOMB as a segment of overlay structure, and this combination of HONEYCOMB and FDCAL-3 forms the detailed calculation subsystem in the JOYPAC system. The detailed calculation subsystem produces the data file of the detailed fundamental informations such as distributions of neutron flux, power etc. about the reactor under stationary performance. This file is required by the quick and simple calculation subsystem SMART and the recording subsystem MASTOR described in Part I. Thus, times of resorting to the time-consuming detailed calculation are reduced as far as possible, and supervision of reactor performance is realized in both features of practically sufficient accuracy and reasonable computer cost. (JPN)

  16. Cooperative transparency for nonproliferation. Technology demonstrations at the Joyo test bed for advanced remote monitoring

    Betsill, J. David; Hashimoto, Yu

    2009-01-01

    The term 'Transparency' has been used widely by many authors and practitioners for various purposes, and there is an assortment of definitions for the term. These definitions vary depending on the field in which the term is used and within the context of its usage. For the purposes of our current project on regional, cooperative nonproliferation transparency and remote monitoring, the relevant field is nuclear nonproliferation, and in this context, we define the term Cooperative Nonproliferation Transparency as: 'Providing sufficient and appropriate information to a cooperating party so that they can independently develop their own evaluation and assessment of the reviewed party regarding their consistency with nonproliferation goals.' Key aspects of cooperative nonproliferation transparency activities include mutually agreeing upon the type of information or data that will be shared, how it will be collected, and who has access to that information. The Japan Atomic Energy Agency's (JAEA) Nonproliferation Science and Technology Center (NPSTC) has been exploring the possible use, development, and application of methods and technologies for Cooperative Transparency for Nonproliferation to support regional confidence building and cooperation n the peaceful use of nuclear energy throughout the East Asia region. (author)

  17. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    Salam, M. A.

    2013-01-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  18. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    Salam, M. A. [Atomic Energy Research Establishment, Dhaka (Bangladesh)

    2013-07-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  19. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

    2018-01-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

  20. X-ray imaging using digital cameras

    Winch, Nicola M.; Edgar, Andrew

    2012-03-01

    The possibility of using the combination of a computed radiography (storage phosphor) cassette and a semiprofessional grade digital camera for medical or dental radiography is investigated. We compare the performance of (i) a Canon 5D Mk II single lens reflex camera with f1.4 lens and full-frame CMOS array sensor and (ii) a cooled CCD-based camera with a 1/3 frame sensor and the same lens system. Both systems are tested with 240 x 180 mm cassettes which are based on either powdered europium-doped barium fluoride bromide or needle structure europium-doped cesium bromide. The modulation transfer function for both systems has been determined and falls to a value of 0.2 at around 2 lp/mm, and is limited by light scattering of the emitted light from the storage phosphor rather than the optics or sensor pixelation. The modulation transfer function for the CsBr:Eu2+ plate is bimodal, with a high frequency wing which is attributed to the light-guiding behaviour of the needle structure. The detective quantum efficiency has been determined using a radioisotope source and is comparatively low at 0.017 for the CMOS camera and 0.006 for the CCD camera, attributed to the poor light harvesting by the lens. The primary advantages of the method are portability, robustness, digital imaging and low cost; the limitations are the low detective quantum efficiency and hence signal-to-noise ratio for medical doses, and restricted range of plate sizes. Representative images taken with medical doses are shown and illustrate the potential use for portable basic radiography.

  1. Sister Lab Program Prospective Partner Nuclear Profile: Malaysia

    Bissani, M; Tyson, S

    2006-01-01

    The Malaysian Deputy Prime Minister Tun Dr Ismail Abdul Rahman suggested in the early 1970s that Malaysia should have a role in the development of nuclear science and technology for peaceful purposes. Accordingly, the Center for the Application of Nuclear Energy (CRANE) was established, with a focus on the development of a scientific and technical pool critical to a national nuclear power program. The Malaysian Cabinet next established the Tun Ismail Atomic Research Center (TIARC) under the Ministry of Science, Technology and the Environment on 19 September 1972, at a site in Bangi, about 35 km south of Kuala Lampur. On 28 June 1982, the PUSPATI reactor, a 1-MW TRIGA MK-II research reactor, first reached criticality. On 10 August 1994, TIARC was officially renamed as the Malaysian Institute for Nuclear Technology Research (MINT). In addition to radioisotope production and neutron radiography conducted at the PUSPATI research reactor, MINT also supports numerous programs employing nuclear technology for medicine, agriculture and industry, and has been involved in both bilateral and multilateral technical cooperation to extend its capabilities. As an energy exporting country, Malaysia has felt little incentive to develop a nuclear energy program, and high level opposition within the government discouraged it further. A recent statement by Malaysia's Science, Technology and Innovation Minister supported this view, indicating that only a near-catastrophic jump in world oil prices might change the government's view. However, the rate at which Malaysia is using its natural gas and oil reserves is expected to force it to reassess the role of nuclear energy in the near future. In addition, the government does intend to construct a radioactive waste repository to dispose of naturally occurring radioactive materials (extracted during tin mining, in particular). Also, Malaysia's growing economy could encourage expansion in Malaysia's existing nuclear-applications programs

  2. Elementary particles and high energy phenomena. Progress report

    Ford, W.T.

    1984-01-01

    During the past year, the MAC collaboration continued accumulating data at PEP and published several results. Included in the results are the lifetime measurement of b quarks; the observation of electroweak interference effects using Bhabba, μμ, and tau tau scattering, a precise measurement of R, observations of direct photons, and a study of energy-energy correlations in multihadron events. For MkII/SLC, a vertex detector system chamber using proportional drift tubes is under construction.The Tagged Photon Spectrometer group has published results and is preparing a new experiment, E-691, to be run in January, 1985. The published results of E-516 include a study of J/psi photoproduction and a study of D 0 → K - π + π 0 . An analysis of Λ 0 anti Λ 0 , and K 0 inclusive photoproduction has recently been completed and is being prepared for publication. Design work and preparation for prototype development has begun on the end cap drift chambers for SLD. Experiment E-400, Hadronic Charm Production, accumulated data during the first two SAVER runs at Fermilab and is now involved in data analysis. E-687, a photoproduction experiment at Fermilab, has begun construction of an electromagnetic calorimeter to measure the energy and postion photons and electrons. Novel forms of supersymmetry breaking were investigated. The anomaly and its relation to the construction of finite supersymmetric theories was studied. A long project was completed on the relation between the gluon condensate in QCD and on effective gluon mass. A series of papers was published on numerical studies of quantum Hamiltonian field theories. Studies of the confinement-deconfinement phase transition in QCD at high temperature were conducted, with applications both to the early universe and to the phenomenogy of ultrarelativistic nucleus-nucleus collisons

  3. Evaluation of machinability and flexural strength of a novel dental machinable glass-ceramic.

    Qin, Feng; Zheng, Shucan; Luo, Zufeng; Li, Yong; Guo, Ling; Zhao, Yunfeng; Fu, Qiang

    2009-10-01

    To evaluate the machinability and flexural strength of a novel dental machinable glass-ceramic (named PMC), and to compare the machinability property with that of Vita Mark II and human enamel. The raw batch materials were selected and mixed. Four groups of novel glass-ceramics were formed at different nucleation temperatures, and were assigned to Group 1, Group 2, Group 3 and Group 4. The machinability of the four groups of novel glass-ceramics, Vita Mark II ceramic and freshly extracted human premolars were compared by means of drilling depth measurement. A three-point bending test was used to measure the flexural strength of the novel glass-ceramics. The crystalline phases of the group with the best machinability were identified by X-ray diffraction. In terms of the drilling depth, Group 2 of the novel glass-ceramics proves to have the largest drilling depth. There was no statistical difference among Group 1, Group 4 and the natural teeth. The drilling depth of Vita MK II was statistically less than that of Group 1, Group 4 and the natural teeth. Group 3 had the least drilling depth. In respect of the flexural strength, Group 2 exhibited the maximum flexural strength; Group 1 was statistically weaker than Group 2; there was no statistical difference between Group 3 and Group 4, and they were the weakest materials. XRD of Group 2 ceramic showed that a new type of dental machinable glass-ceramic containing calcium-mica had been developed by the present study and was named PMC. PMC is promising for application as a dental machinable ceramic due to its good machinability and relatively high strength.

  4. The cAMP Response Element Binding protein (CREB) is activated by Insulin-like Growth Factor-1 (IGF-1) and regulates myostatin gene expression in skeletal myoblast

    Zuloaga, R.; Fuentes, E.N.; Molina, A.; Valdés, J.A.

    2013-01-01

    Highlights: •IGF-1 induces the activation of CREB via IGF-1R/PI3K/PLC signaling pathway. •Calcium dependent signaling pathways regulate myostatin gene expression. •IGF-1 regulates myostatin gene expression via CREB transcription in skeletal myoblast. -- Abstract: Myostatin, a member of the Transforming Growth Factor beta (TGF-β) superfamily, plays an important role as a negative regulator of skeletal muscle growth and differentiation. We have previously reported that IGF-1 induces a transient myostatin mRNA expression, through the activation of the Nuclear Factor of Activated T cells (NFAT) in an IP 3 /calcium-dependent manner. Here we examined the activation of CREB transcription factor as downstream targets of IGF-1 during myoblast differentiation and its role as a regulator of myostatin gene expression. In cultured skeletal myoblast, IGF-1 induced the phosphorylation and transcriptional activation of CREB via IGF-1 Receptor/Phosphatidylinositol 3-Kinase (PI3K)/Phospholipase C gamma (PLC γ), signaling pathways. Also, IGF-1 induced calcium-dependent molecules such as Calmodulin Kinase II (CaMK II), Extracellular signal-regulated Kinases (ERK), Protein Kinase C (PKC). Additionally, we examined myostatin mRNA levels and myostatin promoter activity in differentiated myoblasts stimulated with IGF-1. We found a significant increase in mRNA contents of myostatin and its reporter activity after treatment with IGF-1. The expression of myostatin in differentiated myoblast was downregulated by the transfection of siRNA–CREB and by pharmacological inhibitors of the signaling pathways involved in CREB activation. By using pharmacological and genetic approaches together these data demonstrate that IGF-1 regulates the myostatin gene expression via CREB transcription factor during muscle cell differentiation

  5. Maritime Aerosol optical properties measured by ship-borne sky radiometer

    Aoki, K.

    2017-12-01

    Maritime aerosols play an important role in the earth climate change. We started the measurements of aerosol optical properties since 1994 by using ship-borne sky radiometer (POM-01 MK-II and III; Prede Co. Ltd., Japan) over the ocean. We report the results of an aerosol optical properties over the ocean by using Research Vessel of the ship-borne sky radiometers. Aerosol optical properties observation were made in MR10-02 to MR16-09 onboard the R/V Mirai, JAMSTEC. The sky radiometer measure the direct and diffuse solar radiance with seven interference filters (0.315, 0.4, 0.5, 0.675, 0.87, 0.94, and 1.02 µm). Observation interval was made every five minutes by once, only in daytime under the clear sky conditions. GPS provides the position with longitude and latitude and heading direction of the vessel, and azimuth and elevation angle of the sun. The aerosol optical properties were computed using the SKYRAD.pack version 4.2. The obtained Aerosol optical properties (Aerosol optical thickness, Ångström exponent, Single scattering albedo, and etc.) and size distribution volume clearly showed spatial and temporal variability over the ocean. Aerosol optical thickness found over the near the coast (Asia and Tropical area) was high and variable. The size distribution volume have peaks at small particles at Asian coast and large particles at Tropical coast area. We provide the information, in this presentation, on the aerosol optical properties measurements with temporal and spatial variability in the Maritime Aerosol. This project is validation satellite of GCOM-C/SGLI, JAXA and other. The GCOM-C satellite scheduled to be launched in 2017 JFY.

  6. Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

    Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

    1975-06-01

    The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

  7. The cAMP Response Element Binding protein (CREB) is activated by Insulin-like Growth Factor-1 (IGF-1) and regulates myostatin gene expression in skeletal myoblast

    Zuloaga, R. [Facultad de Ciencias Biológicas, Universidad Andres Bello, Santiago (Chile); Fuentes, E.N.; Molina, A. [Facultad de Ciencias Biológicas, Universidad Andres Bello, Santiago (Chile); Interdisciplinary Center for Aquaculture Research (INCAR), Víctor Lamas 1290, PO Box 160-C, Concepción (Chile); Valdés, J.A., E-mail: jvaldes@unab.cl [Facultad de Ciencias Biológicas, Universidad Andres Bello, Santiago (Chile); Interdisciplinary Center for Aquaculture Research (INCAR), Víctor Lamas 1290, PO Box 160-C, Concepción (Chile)

    2013-10-18

    Highlights: •IGF-1 induces the activation of CREB via IGF-1R/PI3K/PLC signaling pathway. •Calcium dependent signaling pathways regulate myostatin gene expression. •IGF-1 regulates myostatin gene expression via CREB transcription in skeletal myoblast. -- Abstract: Myostatin, a member of the Transforming Growth Factor beta (TGF-β) superfamily, plays an important role as a negative regulator of skeletal muscle growth and differentiation. We have previously reported that IGF-1 induces a transient myostatin mRNA expression, through the activation of the Nuclear Factor of Activated T cells (NFAT) in an IP{sub 3}/calcium-dependent manner. Here we examined the activation of CREB transcription factor as downstream targets of IGF-1 during myoblast differentiation and its role as a regulator of myostatin gene expression. In cultured skeletal myoblast, IGF-1 induced the phosphorylation and transcriptional activation of CREB via IGF-1 Receptor/Phosphatidylinositol 3-Kinase (PI3K)/Phospholipase C gamma (PLC γ), signaling pathways. Also, IGF-1 induced calcium-dependent molecules such as Calmodulin Kinase II (CaMK II), Extracellular signal-regulated Kinases (ERK), Protein Kinase C (PKC). Additionally, we examined myostatin mRNA levels and myostatin promoter activity in differentiated myoblasts stimulated with IGF-1. We found a significant increase in mRNA contents of myostatin and its reporter activity after treatment with IGF-1. The expression of myostatin in differentiated myoblast was downregulated by the transfection of siRNA–CREB and by pharmacological inhibitors of the signaling pathways involved in CREB activation. By using pharmacological and genetic approaches together these data demonstrate that IGF-1 regulates the myostatin gene expression via CREB transcription factor during muscle cell differentiation.

  8. RFX Transcription Factor DAF-19 Regulates 5-HT and Innate Immune Responses to Pathogenic Bacteria in Caenorhabditis elegans

    Choi, Sunju; Xu, Lu; Sze, Ji Ying

    2013-01-01

    In Caenorhabditis elegans the Toll-interleukin receptor domain adaptor protein TIR-1 via a conserved mitogen-activated protein kinase (MAPK) signaling cascade induces innate immunity and upregulates serotonin (5-HT) biosynthesis gene tph-1 in a pair of ADF chemosensory neurons in response to infection. Here, we identify transcription factors downstream of the TIR-1 signaling pathway. We show that common transcription factors control the innate immunity and 5-HT biosynthesis. We demonstrate that a cysteine to tyrosine substitution in an ARM motif of the HEAT/Arm repeat region of the TIR-1 protein confers TIR-1 hyperactivation, leading to constitutive tph-1 upregulation in the ADF neurons, increased expression of intestinal antimicrobial genes, and enhanced resistance to killing by the human opportunistic pathogen Pseudomonas aeruginosa PA14. A forward genetic screen for suppressors of the hyperactive TIR-1 led to the identification of DAF-19, an ortholog of regulatory factor X (RFX) transcription factors that are required for human adaptive immunity. We show that DAF-19 concerts with ATF-7, a member of the activating transcription factor (ATF)/cAMP response element-binding B (CREB) family of transcription factors, to regulate tph-1 and antimicrobial genes, reminiscent of RFX-CREB interaction in human immune cells. daf-19 mutants display heightened susceptibility to killing by PA14. Remarkably, whereas the TIR-1-MAPK-DAF-19/ATF-7 pathway in the intestinal immunity is regulated by DKF-2/protein kinase D, we found that the regulation of tph-1 expression is independent of DKF-2 but requires UNC-43/Ca2+/calmodulin-dependent protein kinase (CaMK) II. Our results suggest that pathogenic cues trigger a common core-signaling pathway via tissue-specific mechanisms and demonstrate a novel role for RFX factors in neuronal and innate immune responses to infection. PMID:23505381

  9. Chandrayaan-2: India's First Soft-landing Mission to Moon

    Mylswamy, Annadurai; Krishnan, A.; Alex, T. K.; Rama Murali, G. K.

    2012-07-01

    . Mission Elements:, On board segment of Chandrayaan-2 mission consists of a lunar Orbiter and a lunar Lander-Rover. The orbiter for Chandrayaan-2 mission is similar to that of Chandrayaan-1 from structural and propulsion aspects. Based on a study of various mission management and trajectory options, such as, separation of the Lander-Rover module in Earth Parking Orbit (EPO) or in lunar transfer trajectory (LTT) or in lunar polar orbit (LPO), the option of separating of this module at LTT, after required midcourse corrections, was selected as this offers an optimum mass and overall mission management advantage. The orbiter propulsion system will be used to transfer Orbiter-Lander-Rover composite from EPO to LTT. On reaching LTT, the Lander-Rover module will be separated from the orbiter module. The Lander-Rover and Orbiter modules are configured with individual propulsion and housekeeping systems. The indigenously developed Geostationary Satellite Launch Vehicle GSLV (Mk-II) will be used for this mission. The most critical aspect of its feasibility was an accurate evaluation of the scope for taking a 3200kg lift off mass into EPO. A Lander-Rover mass of 1270kg (including the propellant for soft landing) will provide sufficient margin for such a lift off within the capability of flight proven GSLV (Mk-II) for the EPO. Mission Scenario: ,GSLV (Mk-II) will launch the Lunar Orbiter coupled to the Lunar Lander-Rover into EPO (170 x 16980 km) following which the Orbiter will boost the orbit from EPO to LTT where the two modules will be separated. Both of them will make their independent journey towards moon and reach lunar polar orbit independently. The orbiter module will be initially placed in a circular polar orbit (200km) and the Lander-Rover module descends towards the lunar surface. After landing, a motorized rover with robotic arm and scientific instruments would be released on to the lunar surface. Although the exact landing location is yet to be finalized, a high

  10. Conceptual Layout of an AHR Building and Facilities

    Ryu, Jeong-Soo; Chae, Hee-Taek; Park, Cheol

    2007-01-01

    A research reactor has been widely utilized in various fields such as industry, engineering, medicine, life science, environment etc., and now application fields are gradually being expanded together with the development of technology. The utilization of research reactor is related to necessary and essential technologies of IT, NT, BT, ET and ST. KAERI has considerable experiences on research reactor technology through the TRIGA MK II, TRIGA MK III, and the HANARO. They have largely contributed to the development of nuclear technology and lead to the nuclear industry. Through the design, construction, operation and utilization of these research reactors, lots of human resources have been developed and the basic and applied technologies in nuclear research fields have been developed step by step. In particular, the HANARO (High-flux Advanced Neutron Application Reactor) of 30MWth, which began to operate from 1995, is a landmark and a nuclear milestone in Korea. The design, construction and operation of the HANARO has allowed for a large progress in the research reactor technology. The active uses of many experimental facilities in the HANARO have enhanced the advancement of nuclear technology and spread the benefits of nuclear R and D results. KAERI is preparing for another leap, i.e., incorporating the experience and nuclear technology accumulated during the design, construction, operation and utilization of the HANARO. KAERI also has human resources who have comprehensive experiences in the design, construction, commissioning, operation and utilization of a research reactor through the HANARO project. Hence, based on such considerable experience and human resource, an advanced HANARO research reactor (AHR) is being developed by KAERI for the future needs of research reactors. Its overall concept is basically similar with the HANARO, but a much improved one. The AHR will be a light water cooled and heavy water moderated and reflected pool type research reactor with

  11. Cyclin D2 is a critical mediator of exercise-induced cardiac hypertrophy.

    Luckey, Stephen W; Haines, Chris D; Konhilas, John P; Luczak, Elizabeth D; Messmer-Kratzsch, Antke; Leinwand, Leslie A

    2017-12-01

    A number of signaling pathways underlying pathological cardiac hypertrophy have been identified. However, few studies have probed the functional significance of these signaling pathways in the context of exercise or physiological pathways. Exercise studies were performed on females from six different genetic mouse models that have been shown to exhibit alterations in pathological cardiac adaptation and hypertrophy. These include mice expressing constitutively active glycogen synthase kinase-3β (GSK-3βS9A), an inhibitor of CaMK II (AC3-I), both GSK-3βS9A and AC3-I (GSK-3βS9A/AC3-I), constitutively active Akt (myrAkt), mice deficient in MAPK/ERK kinase kinase-1 (MEKK1 -/- ), and mice deficient in cyclin D2 (cyclin D2 -/- ). Voluntary wheel running performance was similar to NTG littermates for five of the mouse lines. Exercise induced significant cardiac growth in all mouse models except the cyclin D2 -/- mice. Cardiac function was not impacted in the cyclin D2 -/- mice and studies using a phospho-antibody array identified six proteins with increased phosphorylation (greater than 150%) and nine proteins with decreased phosphorylation (greater than 33% decrease) in the hearts of exercised cyclin D2 -/- mice compared to exercised NTG littermate controls. Our results demonstrate that unlike the other hypertrophic signaling molecules tested here, cyclin D2 is an important regulator of both pathologic and physiological hypertrophy. Impact statement This research is relevant as the hypertrophic signaling pathways tested here have only been characterized for their role in pathological hypertrophy, and not in the context of exercise or physiological hypertrophy. By using the same transgenic mouse lines utilized in previous studies, our findings provide a novel and important understanding for the role of these signaling pathways in physiological hypertrophy. We found that alterations in the signaling pathways tested here had no impact on exercise performance. Exercise

  12. Influence of Flavonoids on Mechanism of Modulation of Insulin Secretion.

    Soares, Juliana Mikaelly Dias; Pereira Leal, Ana Ediléia Barbosa; Silva, Juliane Cabral; Almeida, Jackson R G S; de Oliveira, Helinando Pequeno

    2017-01-01

    The development of alternatives for insulin secretion control in vivo or in vitro represents an important aspect to be investigated. In this direction, natural products have been progressively explored with this aim. In particular, flavonoids are potential candidates to act as insulin secretagogue. To study the influence of flavonoid on overall modulation mechanisms of insulin secretion. The research was conducted in the following databases and platforms: PubMed, Scopus, ISI Web of Knowledge, SciELO, LILACS, and ScienceDirect, and the MeSH terms used for the search were flavonoids, flavones, islets of Langerhans, and insulin-secreting cells. Twelve articles were included and represent the basis of discussion on mechanisms of insulin secretion of flavonoids. Papers in ISI Web of Knowledge were in number of 1, Scopus 44, PubMed 264, ScienceDirect 511, and no papers from LILACS and SciELO databases. According to the literature, the majority of flavonoid subclasses can modulate insulin secretion through several pathways, in an indication that corresponding molecule is a potential candidate for active materials to be applied in the treatment of diabetes. The action of natural products on insulin secretion represents an important investigation topic due to their importance in the diabetes controlIn addition to their typical antioxidant properties, flavonoids contribute to the insulin secretionThe modulation of insulin secretion is induced by flavonoids according to different mechanisms. Abbreviations used: K ATP channels: ATP-sensitive K + channels, GLUT4: Glucose transporter 4, ERK1/2: Extracellular signal-regulated protein kinases 1 and 2, L-VDCCs: L-type voltage-dependent Ca +2 channels, GLUT1: Glucose transporter 1, AMPK: Adenosine monophosphate-activated protein kinase, PTP1B: Protein tyrosine phosphatase 1B, GLUT2: Glucose transporter 2, cAMP: Cyclic adenosine monophosphate, PKA: Protein kinase A, PTK: Protein tyrosine kinase, CaMK II: Ca 2+ /calmodulin

  13. Sister Lab Program Prospective Partner Nuclear Profile: Vietnam

    Bissani, M; Tyson, S

    2006-01-01

    Vietnam's nuclear program began in the 1960s with the installation at Dalat of a 250 kW TRIGA Mk-II research reactor under the U.S. Atoms for Peace Program. The reactor was shut down and its core removed only a few years later, and the nuclear research program was suspended until after the end of the civil war in the late 1970s. The Soviet Union assisted Vietnam in restoring the Dalat reactor to an operational status in 1984, trained a cadre of scientific and technical staff in its operation, and contributed to the development of nuclear science for the medical and agricultural sectors. In the agricultural area in particular, Vietnamese experts have been very successful in developing mutant strains of rice, and continue to work with the IAEA to yield strains that have a shorter growing period, increased resistance to disease, and other desirable characteristics. Rice has always been the main crop in Vietnam, but technical cooperation with the IAEA and other states has enabled the country to become one of the top rice producers in the world, exporting much of its annual crop to over two dozen countries annually. More recently, Vietnam's government has shown increasing interest in developing a civil nuclear program to supplement its fossil fuel and other energy resources. Projections from a variety of open sources, ranging from the IAEA, the U.S. Department of Energy's Energy Information Administration (EIA), the Vietnamese government, energy corporations, and think tanks all predict a massive increase in energy consumption--especially electricity--within Vietnam and the region as a whole. This growth in consumption will require a corresponding increase in energy production, which in Vietnam is currently satisfied mainly by fossil fuels (coal) and renewable energy (hydropower and biomass); Vietnam has a refining capacity of about 800 barrels/day. Most of its crude oil is exported to generate export income, and is not used to generate electricity. Although Vietnam is

  14. The use of neutron activation analysis in environmental pollution studies

    Yusof, A.M.; Gill, S.K.; Salleh, S. [University Technology Malaysia, Dept. of Chemistry, Johor (Malaysia); Akyil, S. [Ege University, Institute of Nuclear Sciences, Izmir (Turkey); Hamzah, S.; Rahman, S.A.; Wood, A.K.H. [Malaysia Institute of Nuclear Technology Research, PUSPATI Complex, Bangi, Selangor (Malaysia)

    2001-11-01

    Environmental samples and localized species from a marine environment, water samples for public drinking, sediment core samples from a polluted marine environment, soil samples from tin-tailing dump sites, air particulate matter and leachates from landfills were analyzed for their trace, toxic elemental contents, chemical species and natural radioactivity in an attempt to assess the safety levels of these pollutants in these matrices by means of instrumental neutron activation analysis (INAA) and other related nuclear techniques. Complementary techniques such as the graphite furnace atomic absorption spectrometry (GFAAS), ICP-MS, ion chromatography and pre-concentration steps particularly in the speciation studies were also incorporated in these studies for specific elemental determinations prior to irradiation in a neutron flux of about 5.1 x 10{sup 8} n.m{sup -2}.s{sup -1} from a TRIGA Mk.II reactor. Pre-concentration of the chemical species of As and Se was done using a mixture of ammonium pyrrolidinethiocarbamatechloroform (APDTC-CHCl{sub 3}) while activated carbon derived from agricultural wastes was used in the iodine speciation. Some of the specific chemical species have to be separated prior to the final quantitative determination to reduce interference and enhance the sensitivity of the INAA technique. These include arsenic, selenium and iodine species present in various matrices. The more toxic inorganic arsenic, selenium, iodine and a host of other trace elements were detected in these samples by quantifying their respective {gamma}-rays emitted from the radioisotopes. The amounts of As(III) present vary from about 1.8 ng/g to 15.5 ng/g in localized marine species, 0.1 ng/g to more than 5.0 ng/g in treated public drinking water while the more toxic inorganic Se (IV) is present in the range of 1.5 {mu}g/L to about 4.5 {mu}g/L. The distribution patterns of pollutants were presented on maps and deductions were made from these patterns to address pollution

  15. Development of early core anomaly detection system by using in-sodium microphone in JOYO. Fundamental characteristics test of in-sodium microphone in water and examination of improvement of detection accuracy

    Komai, Masafumi

    2001-07-01

    Fast reactor core anomalies can be detected in near real-time with acoustic sensors. An acoustic detection system senses an in-core anomaly immediately from the fast acoustic signals that propagate through the sodium coolant. One example of a detectable anomaly is sodium boiling due to local blockage in a sub-assembly; the slight change in background acoustic signals can be detected. A key advantage of the acoustic detector is that it can be located outside the core. The location of the anomaly in the core can be determined by correlating multiple acoustic signals. This report describes the testing and fundamental characteristics of a microphone suitable for use in the sodium coolant and examines methods to improve the system's S/N ratio. Testing in water confirmed that the in-sodium microphone has good impulse and wide band frequency responses. These tests used impulse and white noise signals that imitate acoustic signals from boiling sodium. Correlation processing of multiple microphone signals to improve S/N ratio is also described. (author)

  16. Remote monitoring demonstration

    Caskey, Susan; Olsen, John

    2006-01-01

    The recently upgraded remote monitoring system at the Joyo Experimental Reactor uses a DCM-14 camera module and GEMINI software. The final data is compatible both with the IAEA-approved GARS review software and the ALIS software that was used for this demonstration. Features of the remote monitoring upgrade emphasized compatibility with IAEA practice. This presentation gives particular attention to the selection process for meeting network security considerations at the O'arai site. The Joyo system is different from the NNCA's ACPF system, in that it emphasizes use of IAEA standard camera technology and data acquisition and transmission software. In the demonstration itself, a temporary virtual private network (VPN) between the meeting room and the server at Sandia in Albuquerque allowed attendees to observe data stored from routine transmissions from the Joyo Fresh Fuel Storage to Sandia. Image files from a fuel movement earlier in the month showed Joyo workers and IAEA inspectors carrying out a transfer. (author)

  17. The orange-brown patina of Salisbury Cathedral (West Porch) surfaces: evidence of its man-made origin.

    Martín-Gil, Jesus; Martín-Gil, Francisco Javier; del Carmen Ramos-Sánchez, Maria; Martín-Ramos, Pablo

    2005-09-01

    In this paper, we attempt to elucidate the composition and origin of the orange patina on the surfaces of the West-Porch of Salisbury Cathedral by comparison to other known patinas: (i) the orange-brown patina on the marble surfaces of the Acropolis in Athens and the Arch of Titus in Rome whose analyses have shown very high amounts of phosphates, and generally amino acids from animal-skin glue or other protein binders; (ii) the phosphated patinas which also contain oxalates, found in 1996 on Catalonian calcareous sandstones and in the calcareous dolomites of the Monastery of Silos, Spain, whose origin is either the application of calcium caseinate, or egg yolk and animal glue; and (iii) the patinas with only oxalates found in some of Verona's monuments (St. Zeno) and Spanish sites as in the Monastery of Guadalupe and Cuenca cathedral, formed either by the mineralization of algal filaments or by biological reactions yielding oxalate from yolk egg (added to stone as part of preservative empirical treatments). In the winter of 2003, the West-Porch of Salisbury Cathedral received conservation works, but the old patina was not entirely removed. This fact has allowed us to collect the samples for its study. The IR spectra were registered with a Golden Gate ATR Mk II system using attenuated total reflectance Fourier transform infrared (ATR/FTIR) spectrometry. Mineral composition was determined by XRD (Philips PW 1710 spectrometer with Cu tube), whereas major and trace elements analyses were performed by XRF (Philips PW1480 PW). Microscopy examination was performed on a Leica M655 microscope. Phosphate, oxalate, calcium and sulphate contents were analysed by usual chemical methods. ATD-FTIR spectra of the Salisbury's patina exhibit peaks at 2361, 2341 and 671 cm(-1) (assigned to phosphates); 3410, 1680, 1620, 1122 and 602 cm(-1) (assigned to sulphates); and 1447/1437 and 876 cm(-1) (attributed to carbonates). The little peaks at 1620 and 798 cm(-1) could be assigned to

  18. Actual status of sodium cavitation studies in Japan

    Yamamoto, K.; Kamiyama, S.; Hashimoto, H.; Mochizuki, K.; Nakai, Y.; Ishibashi, E.; Tamaoki, T.

    1976-01-01

    A cavitation test has been conducted on some components of the fast experimental reactor JOYO. Design is in progress for the fast proto-type reactor MONJU. Deliberate consideration has been taken against cavitation as this reactor will be operated under severer service conditions than that of JOYO. A cavitation test of entrance nozzles of MONJU fuel subassemblies was performed in water. In order to obtain design data a program of cavitation tests is planned

  19. Experience on sodium removal from various components

    Kamei, M; Kanbe, M; Yagisawa, H; Sasaki, S; Kataoka, H; Fukada, T; Ishii, Y; Saito, R; Mimoto, Y [O-arai Engineering Centre, PNC, Ibaraki-ken, Tokio (Japan)

    1978-08-01

    Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

  20. Experience on sodium removal from various components

    Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.; Fukada, T.; Ishii, Y.; Saito, R.; Mimoto, Y.

    1978-01-01

    Since 1970, OEC (O-arai Engineering Center) has been Investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of 'JOYO' and Dummy fuel assembly of 'JOYO', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of 'JOYO', a sector model of Sodium-to-Air cooler of 'JOYO' and a proto-type isolation valve of 'JOYO' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental sub-assemblies, the Fuel Handling Machine of 'MONJU' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a 'JOYO' prototype pump by reinstalling it after sodium removal five times. (author)

  1. Experience on sodium removal from various components

    Kamei, M.; Kanbe, M.; Yagisawa, H.; Sasaki, S.; Kataoka, H.

    1978-02-01

    Since 1970, OEC (O-arai Engineering Center) has been investigating the following methods for removal of sodium from the components of sodium plants: steam cleaning for the 50 MW Steam Generator, secondary proto-type pump of ''JOYO'' and Dummy fuel assembly of ''JOYO'', alcohol cleaning for Sector Model of Intermediate Heat Exchanger (IHX) of ''JOYO'', a sector model of Sodium-to-Air cooler of ''JOYO'' and a proto-type Isolation valve of ''JOYO'' and cleaning by vacuumization at high temperature for Regenerative Heat Exchanger. This report describes the outline of the Sodium Disposal Facility and experience of sodium removal processing on the 50 MW Steam Generator, the crevices of the experimental subassemblies, the Fuel Handling Machine of ''MONJU'' and the Regenerative Heat Exchanger of the Sodium Flow Test Facility. Through these experiences it was noted that, (1) Removal of Sodium from crevices such as in bolted joints are very difficult. (2) Consideration is needed in the removal process where material damage might occur from the generation of hydro-oxides. (3) Some detection device to tell the completion of sodium removal as well as the end of reaction is required. (4) Requalification rules should be clarified. Efforts in this direction have been made in the case of a ''JOYO'' prototype pump by reinstalling it after sodium removal five times. (author)

  2. Remote monitoring technologies and applications. JAEA-SNL technical cooperation experience in RM for nuclear transparency

    Matter, John

    2006-01-01

    In ten years of remote monitoring cooperation, Sandia National Laboratories (SNL) and the JAEA (formerly JNC) have developed technology and demonstrated it at the Joyo Experimental Reactor. The program goals were to develop technology to support international safeguards, help evaluate and standardize the technologies for safeguards uses, and demonstrate them for potential regional cooperation. This paper described three generations of remote monitoring systems at the Joyo Fresh Storage and at one of the Joyo Spent Fuel Ponds. Communications and control methods within the facility and between the facility and the remote viewer have changed rapidly. The current configuration is similar to an international safeguards installation, but provides a foundation for transparency cooperation between the JAEA and SNL. Plans to expand this cooperation to other partners are noted. (author)

  3. Evaluation of fast experimental reactor claddings, (2)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro; Tanaka, Yasumasa

    1974-01-01

    Thin-walled fine tubes of Type 316 austenitic stainless steel are used for fuel cladding in Joyo (experimental FBR). The material exhibits the change of the mechanical properties in long-time annealing at high temperature, resulting from the precipitation of carbide in structure. In this connection, the experiment and the results on the changes of the microstructure and mechanical properties (proof stress and hardness) are described. The test specimens are the fuel cladding tubes produced for trial for Joyo core and those for FFTF core made in the U.S.A. They were heated between 400 0 and 850 0 C for 1000 hr in vacuum. (Mori, K.)

  4. Progress report on fast breeder reactor development at PNC, Japan, October - December, 1974

    1975-03-01

    Following the completion of building construction and equipment installation for the experimental fast breeder reactor ''Joyo'' at PNC's Oarai Engineering Center, hydrostatic pressure and leak tests were conducted on the reactor vessel. For the prototype fast breeder reactor ''Monju'', specification was finalized after the design adjustment. For the period from October to December, 1974, the following matters are described: construction of the Joyo, design of the Monju, reactor physics, components and equipments, instruments and control, sodium technology, fuel and material research and development, safety research and development, and steam generator. (Mori, K.)

  5. Review of the activities in Japan

    Otake, I.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress through.operation of the experimental fast reactor JOYO, design of the prototype fast breeder reactor MONJU and related R and D works. JOYO began operation in mid-1977, increased power from 50 MWt to 75 MWth in July 1979 and operation cycles at 75 MWth are continued at present. With respect to MONJU, which is a 300 MWe plant, progress toward construction has been made and the safety review are started by the concerned authorities. Conceptual design studies of large demonstration fast breeder reactor are also being made by PNC. It is a 1000 MWe, loop type plant

  6. Dissolution of powdered spent fuel and U crystallization from actual dissolver solution for 'NEXT' process development

    Nomura, Kazunori; Hinai, Hiroshi; Nakahara, Masaumi; Kaji, Naoya; Kamiya, Masayoshi; Ohyama, Koichi; Sano, Yuichi; Washiya, Tadahiro; Komaki, Jun

    2008-01-01

    The beaker-scale experiments on the effective powdered fuel dissolution and the U crystallization from dissolver solution with the irradiated MOX fuel from the experimental fast reactor 'JOYO' were carried out. The powdered fuel was effectively dissolved into the nitric acid solution. In the U crystallization experiments, U crystal was obtained from the actual dissolver solution without any addition of reagent. (authors)

  7. Studies on modeling to failed fuel detection system response in LMFBR

    Miyazawa, T.; Saji, G.; Mitsuzuku, N.; Hikichi, T.; Odo, T.; Rindo, H.

    1981-05-01

    Failed Fuel Detection (FFD) system with Fission Products (FP) detection is considered to be the most promissing method, since FP provides direct information against fuel element failure. For designing FFD system and for evaluating FFD signals, some adequate FFD signal response to fuel failure have been required. But few models are available in nowadays. Thus Power Reactor and Nuclear Fuel Development Corporation (PNC) had developed FFD response model with computer codes, based on several fundamental investigations on FP release and FP behavior, and referred to foreign country experiences on fuel failure. In developing the model, noble gas and halogen FP release and behavior were considered, since FFD system would be composed of both cover gas monitoring and delayed neutron monitoring. The developed model can provide typical fuel failure response and detection limit which depends on various background signals at cover gas monitoring and delayed neutron monitoring. According to the FFD response model, we tried to assume fuel failure response and detection limit at Japan experimental fast reactor ''JOYO''. The detection limit of JOYO FFD system was estimated by measuring the background signals. Followed on the studies, a complete computer code has been now made with some improvement. On the paper, the details of the model, out line of developed computer code, status of JOYO FFD system, and trial assumption of JOYO FFD response and detection limit. (author)

  8. Experience of Integrated Safeguards Approach for Large-scale Hot Cell Laboratory

    Miyaji, N.; Kawakami, Y.; Koizumi, A.; Otsuji, A.; Sasaki, K.

    2010-01-01

    The Japan Atomic Energy Agency (JAEA) has been operating a large-scale hot cell laboratory, the Fuels Monitoring Facility (FMF), located near the experimental fast reactor Joyo at the Oarai Research and Development Center (JNC-2 site). The FMF conducts post irradiation examinations (PIE) of fuel assemblies irradiated in Joyo. The assemblies are disassembled and non-destructive examinations, such as X-ray computed tomography tests, are carried out. Some of the fuel pins are cut into specimens and destructive examinations, such as ceramography and X-ray micro analyses, are performed. Following PIE, the tested material, in the form of a pin or segments, is shipped back to a Joyo spent fuel pond. In some cases, after reassembly of the examined irradiated fuel pins is completed, the fuel assemblies are shipped back to Joyo for further irradiation. For the IAEA to apply the integrated safeguards approach (ISA) to the FMF, a new verification system on material shipping and receiving process between Joyo and the FMF has been established by the IAEA under technical collaboration among the Japan Safeguard Office (JSGO) of MEXT, the Nuclear Material Control Center (NMCC) and the JAEA. The main concept of receipt/shipment verification under the ISA for JNC-2 site is as follows: under the IS, the FMF is treated as a Joyo-associated facility in terms of its safeguards system because it deals with the same spent fuels. Verification of the material shipping and receiving process between Joyo and the FMF can only be applied to the declared transport routes and transport casks. The verification of the nuclear material contained in the cask is performed with the method of gross defect at the time of short notice random interim inspections (RIIs) by measuring the surface neutron dose rate of the cask, filled with water to reduce radiation. The JAEA performed a series of preliminary tests with the IAEA, the JSGO and the NMCC, and confirmed from the standpoint of the operator that this

  9. Watt and joule balances

    Robinson, Ian A.

    2014-04-01

    The time is fast approaching when the SI unit of mass will cease to be based on a single material artefact and will instead be based upon the defined value of a fundamental constant—the Planck constant—h . This change requires that techniques exist both to determine the appropriate value to be assigned to the constant, and to measure mass in terms of the redefined unit. It is important to ensure that these techniques are accurate and reliable to allow full advantage to be taken of the stability and universality provided by the new definition and to guarantee the continuity of the world's mass measurements, which can affect the measurement of many other quantities such as energy and force. Up to now, efforts to provide the basis for such a redefinition of the kilogram were mainly concerned with resolving the discrepancies between individual implementations of the two principal techniques: the x-ray crystal density (XRCD) method [1] and the watt and joule balance methods which are the subject of this special issue. The first three papers report results from the NRC and NIST watt balance groups and the NIM joule balance group. The result from the NRC (formerly the NPL Mk II) watt balance is the first to be reported with a relative standard uncertainty below 2 × 10-8 and the NIST result has a relative standard uncertainty below 5 × 10-8. Both results are shown in figure 1 along with some previous results; the result from the NIM group is not shown on the plot but has a relative uncertainty of 8.9 × 10-6 and is consistent with all the results shown. The Consultative Committee for Mass and Related Quantities (CCM) in its meeting in 2013 produced a resolution [2] which set out the requirements for the number, type and quality of results intended to support the redefinition of the kilogram and required that there should be agreement between them. These results from NRC, NIST and the IAC may be considered to meet these requirements and are likely to be widely debated

  10. Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

    Mizuno, M.; Enokido, Y.; Itaki, T.; Kono, K.; Unno, I.; Yamanouchi, S.

    1985-01-01

    A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis

  11. Research and development of bellows for LMFBR in Japan

    Takahashi, T.; Mukai, K.; Yamamoto, K.

    1980-01-01

    Bellows are employed as useful mechanical elements with their flexibility and imperviousness to liquid and gas in the system in which such chemically active substances as sodium are handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: control rod drive mechanism; intermediate heat exchanger; small valve; mechanical penetration assembly of the containment boundary; outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for FBR use in Japan. (author)

  12. Research and development of bellows for LMFBR in Japan

    Takahashi, Tadao; Mukai, Kazuo; Yamamoto, Ken.

    1979-11-01

    The bellows is employed as a useful mechanical element with its flexibility and imperviousness to liquid and gas in the system in which such chemically active substance as sodium is handled. Since the early time of development of Japanese LMFBR, bellows have been used e.g. for the shaft seal of small sodium valves. Bellows are employed in the fast experimental reactor JOYO which is now in operation and the fast prototype reactor MONJU whose design program is in final stage at the following parts: - control rod drive mechanism, - intermediate heat exchanger, - small valve, - mechanical penetration assembly of the containment boundary, - outer piping of the double-walled primary system (for JOYO only). In addition, the application of bellows as thermal expansion joint to the main piping system is under consideration for future FBRs. This paper outlines the research and development work on bellows for the FBR use in Japan. (author)

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

    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)

  14. Transport of radioactive corrosion products in primary system of sodium-cooled fast breeder reactor 'MONJU'

    Matuo, Youichirou; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya

    2011-01-01

    Radioactive corrosion products (CP) are primary cause of personal radiation exposure during maintenance work at FBR plants with no breached fuel. The PSYCHE code has been developed based on the Solution-Precipitation model for analysis of CP transfer behavior. We predicted and analyzed the CP solution and precipitation behavior of MONJU to evaluate the applicability of the PSYCHE code to MONJU, using the parameters verified in the calculations for JOYO. From the calculation result pertaining to the MONJU system, distribution of 54 Mn deposited in the primary cooling system over 20 years of operation is predicted to be approximately 7 times larger than that of 60 Co. In particular, predictions show a notable tendency for 54 Mn precipitation to be distributed in the primary pump and cold-leg. The calculated distribution of 54 Mn and 60 Co in the primary cooling system of MONJU agreed with tendencies of measured distribution of JOYO. (author)

  15. LMFBR plant parameters

    1979-03-01

    This document contains up-to-date data on existing or firmly decided prototype or demonstration LMFBR reactors (Table I), on planned commercial size LMFBR according to the present status of design (Table II) and on experimental fast reactors such as BOR-60, DFR, EBR-II, FERMI, FFTF, JOYO, KNK-II, PEC, RAPSODIE-FORTISSIMO (Table III). Only corrected and revised parameters submitted by the countries participating in the IWGFR are included in this document

  16. Calibration of a He accumulation fluence monitor for fast reactor dosimetry

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

    1997-03-01

    The helium accumulation fluence monitor (HAFM) has been developed for a fast reactor dosimetry. The HAFM measurement system was calibrated using He gas and He implanted samples and the measurement accuracy was confirmed to be less than 5%. Based on the preliminary irradiation test in JOYO, the measured He in the {sup 10}B type HAFM agreed well with the calculated values using the JENDL-3.2 library. (author)

  17. Activities on Fast Reactor Knowledge Preservation in Japan

    Uto, Nariaki; Ohira, Hiroaki

    2013-01-01

    Japanese FRs are stepping up from Joyo, Monju to JSFR with efforts to overcome the Fukushima no. 1 Nuclear Power Station Accidents, including focus on ensuring and enhancing safety of these FRs. Japan (JAEA) has made continuous and contributive efforts to supply FR information to IAEA/INIS. JAEA has also contributed to a global dissemination of its R&D activities through JOPSS English version, which will be beneficial to worldwide FR development

  18. International experience with the bundle behavior of fuel elements of sodium cooled reactors; derivation of a figure of merit for the judgement of fuel pin bundle parameters with respect to abrasion due to thermoelastic pin-pin interaction

    Toebbe, H.

    1987-10-01

    The report describes the status of experience with respect to the abrasion behavior of bundles in standard fuel elements and test elements with wire or grid spacing in the reactors Rapsodie fortissimo, Phenix, DFR, PFR, EBR-II, FFTF, JOYO and KNK II. With the help of simple considerations concerning thermoelastic pin-pin interactions a figure of merit is deduced from the different bundle parameters, which allows a comparative judgement of the parameters of different bundle concepts [de

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

    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)

  20. Progress report on fast breeder reactor development in Japan

    1976-05-01

    Following the completion of the in-argon, high temperature test, the in-sodium functional test of Joyo has set in. The fabrication of the equipments for monitoring the flow rate and temperature in the center channel and the power distribution was finished. The modification of design of the prototype fast breeder reactor Monju came into its phase 3. The interim report on the check-up and review of the Monju project by the Government is now ready. Various calculation codes were developed or are in development stage. The mock-up assembly FCA 7-1 has been built, which consists of the Pu-fueled sector region simulating the Monju core and the U-235-fueled driver region. Various reactor physics experiments have been carried out in this assembly. Also, the calculation methods for reactor physics parameters have been developed, and the detailed calculation on the main shield of Joyo was performed. The situation of the developments of the components for Joyo and Monju and the measuring and control systems is shown. Almost all the existing sodium test facilities in Oarai Engineering Center were in service without any trouble, and the new test facility named ''Carbon transfer test loop'' was commissioned. The progress in the fields of sodium technology, fuel and material, safety and steam generators is reported. (Kako, I.)

  1. Reduction in degree of absorber-cladding mechanical interaction by shroud tube in control rods for the fast reactor

    Donomae, Takako; Katsuyama, Kozo; Tachi, Yoshiaki; Maeda, Koji; Yamamoto, Masaya; Soga, Tomonori

    2011-01-01

    Research and development of a long-life control rod for fast reactors is being conducted at Joyo. One of the challenges in developing a long-life control rod is the restraint of absorber-cladding mechanical interaction (ACMI). First, a helium-bonding rod was selected as a control rod for the experimental fast reactor Joyo, which is the first liquid metal fast reactor in Japan. Its lifetime was limited by ACMI, which is induced by the swelling and relocation of B 4 C pellets. To restrain ACMI, a shroud tube was inserted into the gap between the B 4 C pellets and the cladding tube. However, once B 4 C pellets cracked and broke into small fragments, relocation occurred. After this, the narrow gap closed immediately as the degree of B 4 C pellet swelling increased. To solve this problem, the gap was widened during design, and sodium was selected as the bonding material instead of helium to restrain the increase in pellet temperature. Irradiation testing of the modified sodium-bonding control rod confirmed that ACMI would be restrained by the shroud tube regardless of the occurrence of B 4 C pellet relocation. As a result of these improvements, the estimated lifetime of the control rod at Joyo was doubled. In this paper, the results of postirradiation examination are reported. (author)

  2. Thermal and irradiation effects on high-temperature mechanical properties of materials for SCWR fuel cladding

    Kano, F.; Tsuchiya, Y.; Oka, K.

    2009-01-01

    The thermal and irradiation effects on high-temperature mechanical properties are examined for candidate alloys for fuel cladding of supercritical water-cooled reactors (SCRWs). JMTR (Japan Materials Testing Reactor) and Experimental Fast Reactor JOYO were utilized for neutron irradiation tests, considering their fluence and temperature. Irradiation was performed with JMTR at 600degC up to 4x10 24 n/m 2 and with JOYO at 600degC and 700degC up to 6x10 25 n/m 2 . Tensile test, creep test and hardness measurement were carried out for high-temperature mechanical properties. Based on the uniaxial creep test, the extrapolation curves were drawn with time-temperature relationships utilizing the Larson and Miller Parameter. Several candidate alloys are expected to satisfy the design requirement from the estimation of the creep rupture stress for 50000 hours. Comparing the creep strengths under irradiated and unirradiated conditions, it was inferred that creep deformation was dominated by the thermal effect rather than the irradiation at SCWR core condition. The microstructure was examined using transmission electron microscope (TEM) analysis, focusing on void swelling and helium (He) bubble formation. Void formation was observed in the materials irradiated with JOYO at 600degC but not at 700degC. However, its effect on the deformation of components was estimated to be tolerable since their size and density were negligibly small. The manufacturability of the thin-wall, small-diameter tube was confirmed for the potential candidate alloys through the trial tests in the factory where the fuel cladding tube is manufactured. (author)

  3. Preparation of computer codes for analyzing sensitivity coefficients of burnup characteristics (2) (Contract research, translated document)

    Hanaki, Hiroshi; Sanda, Toshio; Ohashi, Masahisa

    2008-10-01

    To develop nuclear design of LMFBR cores, they are important subjects of research and development to improve the accuracy in nuclear design of large LMFBR cores and to design highly efficient core more rationally. The adjusted nuclear cross-sections library has been made by being reflected the result of critical experiment of the JUPITER, etc. effectively as much as possible. And the distinct improvement of the accuracy in nuclear design of large LMFBR cores has been achieved. In the design of large LMFBR cores, however, it is important to accurately estimate not only nuclear characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. Therefore, it is thought to improve the prediction accuracy for burnup characteristics using many burnup data of 'Joyo' effectively. It is thought the best way to adjust cross sections using sensitivity coefficients of burnup characteristics to utilize burnup data of 'Joyo'. It is able to know the accuracy quantitatively for burnup characteristics of large LMFBR by analyzing the sensitivity coefficients. Therefore in this work computer codes for analyzing sensitivity coefficients of burnup characteristics had been prepared since 1992. In 1992 cross-section adjustment was done by using the data of 'Joyo' and the effect was studied. In this year the adequacy of the codes was studied with a view of applying of design of large LMFBR cores. The results are as follows: (1) The computer codes which could analyze sensitivity coefficients of burnup characteristics taking into consideration plural cycles and refueling were prepared, therefore it came of be able to adjust cross sections using burnup data and to estimate the accuracy for design of large LMFBR cores. The characteristics are not only burnup reactivity loss, breeding ratio but also number density, criticality, reactivity worth, reaction rate ratio, and reaction rate

  4. Argon cover gas purity control on LMFBR

    Maeda, Hiroshi; Kobayashi, Takayoshi [PNC (Japan); Ishiyama, Satoshi [Toshiba (Japan); Motonaga, Tetsuji [Hitachi (Japan)

    1987-07-01

    Various control methods on chemical impurities and radioactive materials (fission products) in the primary argon gas of LMFBRs' have been studied based on experiences in Joyo and results of research and development. These results are reflected on MONJU design. On-line gas chromatographs are installed both in the Primary and in the Secondary Argon Gas Systems in JOYO. Also, chemical analysis has been done by batch sampling in JOYO. Though the rise of impurity concentration had been measured after periodical fuel exchange operation, impurity concentration has been controlled sufficiently under target control limits. In MONJU detailed design, the Rare Gas Removal and Recovery System which consisted of cryogenic distillation equipment had been eliminated and the capacity of Charcoal Beds in the Primary Argon Gas System has been improved to keep the concentration of radioactive materials sufficient low levels. The necessity to control the impurities in fresh argon gas which is supplied to the Primary Argon Gas System is now considered to keep the concentration of Kr and Xe isotopes in specified level, because their isotopes may make background rise for the Tagging Gas Failed Fuel Detection and Location System. Based on various investigations performed on sodium vapor trapping to obtain its detailed characteristics, design specifications and operating conditions of MONJU's Vapor Traps have been decided. To keep the level of radioactivity in gaseous effluents to the environment as low as reasonably achievable, the following means are now adopted in MONJU: the Primary Argon Gas System is composed of a closed recirculating path, but the exhaust gas discharged has different path after the Charcoal Beds; fresh argon gas is blown down to prevent Primary Argon Gas from releasing to the circumference during opening of the primary argon gas boundary, such as fuel exchange operations. (author)

  5. Integral test on activation cross section of tag gas nuclides using fast neutron spectrum fields

    Aoyama, Takafumi; Suzuki, Soju [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-03-01

    Activation cross sections of tag gas nuclides, which will be used for the failed fuel detection and location in FBR plants, were evaluated by the irradiation tests in the fast neutron spectrum fields in JOYO and YAYOI. The comparison of their measured radioactivities and the calculated values using the JENDL-3.2 cross section set showed that the C/E values ranged from 0.8 to 2.8 for the calibration tests in YAYOI and that the present accuracies of these cross sections were confirmed. (author)

  6. Upgrade of reactor operation technology

    Itoh, Hideaki; Suzuki, Toshiaki; O-kawa, Toshikatsu

    2003-01-01

    To improve operational reliability and availability, the operation technology for a fast reactor was developed in the ''JOYO''. This report describes the upgrading of the simulator, plant operation management tools and fuel handling system for the MK-III core operation. The simulator was modified to the MK-III version to verify operation manuals, and to train operators in MK-III operation. The plant operation management tool was replaced on the operation experience to increase the reliability and efficiency of plant management works relating to plant operation and maintenance. To shorten the refueling period, the fuel handling system was upgraded to full automatic remote control. (author)

  7. Automatization of the radiation control measurements

    Seki, Akio; Ogata, Harumi; Horikoshi, Yoshinori; Shirai, Kenji

    1988-01-01

    Plutonium Fuel Production Facility (PFPF) was constructed to fabricate the MOX fuels for 'MONJU' and 'JOYO' reactors and to develop the practical fuel fabricating technology. For the fuel fabrication process in this facility, centralized controlling system is being adopted for the mass production of the fuel and reduction of the radiation exposure dose. Also, the radiation control systems are suitable for the large-scale facility and the automatic-remote process of the fuel fabrication. One of the typical radiation control systems is the self moving survey system which has been developed by PNC and adopted for the automatic routine monitoring. (author)

  8. Status of LMFBR development project in Japan

    Nagane, G.; Akebi, M.; Matsuno, Y.

    1987-01-01

    Initiation of the LMFBR development project in Japan was decided by the Atomic Energy Commission of Japan in 1966. In 1967, the Power Reactor and Nuclear Fuel Development Corporation (PNC) was established to realize the project as a part of its tasks of a wide scope covering all the reseatch and development activities concerning fuel cycle. In the present paper the status of experimental fast reactor (Joyo), which is the first milestone of the LMFBR project, prototype fast reactor (Monju) and R and D activities supporting the project including that for larger LMFBRs in the future is described. (author)

  9. Development of the alcohol waste processing equipment

    Obara, Kiyoshi; Ooyama, Etsuo; Suzuki, Toshiaki; Oohara, Norikazu

    2004-01-01

    In the experimental fast Reactor JOYO, gripper of Fuel Handling Machine and Ex-Vessel Transfer Machine that the sodium adhered is being washed with alcohol. This radioactive alcohol waste that was used to the washing is stored to the tank. If it is able to separate the alcohol and sodium in the alcohol waste it becomes possible to dispose of the alcohol waste. Japan Nuclear Institute and Fuji Electric Systems CO., LTD. Developed the device that adds carbonic acid gas to the alcohol waste and cause the sodium in the alcohol waste separated as carbonate and remove this carbonate by using the thin film evaporator. (author)

  10. A review of fast reactor progress in Japan

    Tomabechi, K [Power Reactor and Nuclear Fuel Development Corporation, Tokyo (Japan)

    1978-07-01

    The fast reactor development project in Japan is continuing at a slightly increased scale of effort in budget. The total budget for LMFBR development for fiscal year 1978 was 24 billion yen. In August 1977 major industries engaged in LMFBR have set up an office where design work can be jointly conducted. Highlights and topics of the fast reactor development activities cover description of JOYO reactor, its first criticality experiment, and the prototype fast breeder MONJU. Research and development programmes dealt with fission products release and its possible interaction with the soodium coolant, inspection of reactor components, experiments simulating sodium leakage, development of steam generator.

  11. Fission and corrosion products behavior in primary circuits of LMFBR's

    Feuerstein, H.; Thorley, A.W.

    1987-08-01

    Most of the 20 presented papers report items belonging to more than one session. The equipment results of primary circuits of LMFBR's relative to corrosion and fission products, release and chemistry of fuel, measurement techniques and analytical procedures of sodium sampling, difficulties with radionuclides and particles, reactor experiences with EBR-II, FFTF, BR10, BOR60, BN350, BN600, JOYO, and KNK-II, DFR, PFR, RAPSODIE, PHENIX, and SUPERPHENIX, and at least the verification of codes for calculation models of radioactive products accumulation and distribution are described. All 20 papers presented at the meeting are separately indexed in the database. (DG)

  12. Sodium vapour deposition studies related to the development of components for LMFBR in Japan

    Himeno, Y; Yamamoto, K; Takahashi, J; Mochizuki, K; Saito, R [O-arai Engineering Centre, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai Machi, Ibaraki-ken (Japan)

    1977-01-01

    During the course of research and development for the experimental fast breeder reactor JOYO, performance tests for such components as rotating plug and vapor trap have been carried out using prototype models. In recent years, as the development effort has been concentrated to the prototype fast reactor MONJU, performance tests for the components mentioned above in large scale are now in progress. In this report, brief description of the test results and experiences obtained from these studies are presented, with the results of basic studies related to these. (author)

  13. Research in decommissioning techniques for nuclear fuel cycle facilities in JNC. 7. JWTF decommissioning techniques

    Ogawa, Ryuichiro; Ishijima, Noboru

    1999-02-01

    Decommissioning techniques such as radiation measuring and monitoring, decontamination, dismantling and remote handling in the world were surveyed to upgrading technical know-how database for decommissioning of Joyo Waste Treatment Facility (JWTF). As the result, five literatures for measuring and monitoring techniques, 14 for decontamination and 22 for dismantling feasible for JWTF decommissioning were obtained and were summarized in tables. On the basis of the research, practical applicability of those techniques to decommissioning of JWTF was evaluated. This report contains brief surveyed summaries related to JWTF decommissioning. (H. Itami)

  14. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    MH Lane

    2006-01-01

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations

  15. Some study on radiation resistance and reliability of piston ring of waste gas compressor for fast breeder experimental reactor

    Muramatsu, Takio; Hidaka, Tsukasa

    1976-01-01

    In the fast breeder experimental reactor ''Joyo'', the gaseous wastes such as reactor cover argon, reactor seal nitrogen gas, fuel handling waste gas etc. shall be collected, compressed and storaged for decaying their activity. Compressors applied in the above process have new type oilless piston rings of Teflon filled with graphite, which might be affected by radioactivity of the waste gases. This report deals with some study on the gamma iradiation effects on the plastic piston rings such as tensile strength, elongation, shock and hardness effects under several irradiation doses and on durability test of the irradiated piston rings under the same compression ratio. (auth.)

  16. Development of fabrication method for thermal expansion difference irradiation temperature monitor

    Noguchi, Kouichi; Takatsudo, Hiroshi; Miyakawa, Shun-ichi; Kobori, Takahisa; Miyo, Toshimasa

    1998-03-01

    This report describes the development activities for the fabrication of the Thermal Expansion Difference irradiation temperature monitor (TED) at the Oarai Engineering Center (OEC)/PNC. TED is used for various irradiation tests in the experimental fast reactor JOYO. TED is the most accurate off-line temperature monitor used for irradiation examination. The TED is composed of a metallic sphere lid and either a stainless steel or nickel alloy container. Once the container is filled with sodium, the metallic sphere lid is sealed by using a resistance weld. This capsule is then loaded into a reactor. Once a TED is loaded into the JOYO reactor, the sodium inside the metallic container increases as a result of thermal expansion. The TED identifies the peak irradiation temperature of the reactor based on a formula correlating temperature to increment values. This formula is established specifically for the particular TED being used during a calibration process performed when the TED is fabricated. Initially the TED was developed by Argonne National Laboratory (ANL) in the United States, and was imported by PNC for use in the JOYO reactor. In 1992 PNC decided to fabricate TED domestically in order to ensure the stability of future supplies. Based on technical information provided by ANL, PNC began fabrication of a TED on an experimental basis. In addition, PNC endeavored to make the domestically produced TED more efficient. This involved improving the techniques used in the sodium filling and the metallic sphere welding processes. These quality control efforts led to PNC's development of processes enabling the capsules to be filled with sodium to nearly 100%. As a result, the accuracy of the temperature dispersion in the out-pile calibration test was improved from +/-10degC to +/-5degC. In 1996 the new domestically fabricated TED was attached to a JOYO irradiation rig. In March of 1997, irradiation of the rig was started on the 30th duty cycle operation, and should be

  17. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  18. The development of breeder reactors in Japan

    Segawa, M.

    1984-01-01

    In the framework of a global analysis of the various available sources of energy, Japan has reserved a prominent place to the nuclear energy and, in the long-term view, to the breeder reactor which will be due for commercial deployment in 20)10. To achieve these objectives, three stages are envisaged, one of the experimental reactor Joyo (in service), one of the demonstration reactor Monju (its construction has been decided), and one of the pre-commercial reactor (due to be taken in hand at the beginning of the Nineties). Efforts will be made in parallel concerning the fuel cycle [fr

  19. 巨椋低平流域の都市化と内水(11)-内水排除施設計画の最適化-

    角屋, 睦; 近森, 秀高

    1992-01-01

    The Ogura low-lying basin located in the south of Kyoto has been urbanized rapidly in recent years. To cope with changes of flood runoff due to urbanization, the improvement of the River Furu was begun in 1971, and the Kumiyama Pump station with a pump of 30 m^3/s was constructed at the downstream end of the River Furu in 1973. Moreover, a pump of 30 m^3/s was added to this pump station in 1987, and the Joyo Pump station with two pumps of 5 m^3/s was constructed at Haccho located in the upstr...

  20. Training in remote monitoring technology. Digital camera module-14(DCM-14)

    Caskey, Susan

    2006-01-01

    The DCM-14 (Digital Camera Module) is the backbone of current IAEA remote monitoring surveillance systems. The control module is programmable with features for encryption, authentication, image compression and scene change detection. It can take periodic or triggered images under a variety of time sequences. This training session covered the DCM-14 features and related programming in DCMSET. It also described the processes for receiving, archiving and backing up the camera images using DCMPOLL and GEMINI software. Setting up a DCM-14 camera controller in the configuration of the remote monitoring system at Joyo formed an exercise. (author)

  1. Transfer of test samples and wastes between post-irradiation test facilities (FMF, AGF, MMF)

    Ishida, Yasukazu; Suzuki, Kazuhisa; Ebihara, Hikoe; Matsushima, Yasuyoshi; Kashiwabara, Hidechiyo

    1975-02-01

    Wide review is given on the problems associated with the transfer of test samples and wastes between post-irradiation test facilities, FMF (Fuel Monitoring Facility), AGF (Alpha Gamma Facility), and MMF (Material Monitoring Facility) at the Oarai Engineering Center, PNC. The test facilities are connected with the JOYO plant, an experimental fast reactor being constructed at Oarai. As introductory remarks, some special features of transferring irradiated materials are described. In the second part, problems on the management of nuclear materials and radio isotopes are described item by item. In the third part, the specific materials that are envisaged to be transported between JOYO and the test facilities are listed together with their geometrical shapes, dimensions, etc. In the fourth part, various routes and methods of transportation are explained with many block charts and figures. Brief explanation with lists and drawings is also given to transportation casks and vessels. Finally, some future problems are discussed, such as the prevention of diffusive contamination, ease of decontamination, and the identification of test samples. (Aoki, K.)

  2. Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

    Tomabechi, K.

    1975-01-01

    The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

  3. Accountability control system in plutonium fuel facility

    Naruki, Kaoru; Aoki, Minoru; Mizuno, Ohichi; Mishima, Tsuyoshi

    1979-01-01

    More than 30 tons of plutonium-uranium mixed-oxide fuel have been manufactured at the Plutonium Facility in PNC for JOYO, FUGEN and DCA (Deuterium Critical Assembly) and for the purpose of irradiation tests. This report reviews the nuclear material accountability control system adopted in the Plutonium Facility. Initially, the main objective of the system was the criticality control of fissible materials at various stages of fuel manufacturing. The first part of this report describes the functions and the structure of the control system. A flow chart is provided to show the various stages of material flow and their associated computer files. The system is composed of the following three sub-systems: procedures of nuclear material transfer; PIT (Physical Inventory Taking); data retrieval, report preparation and file maintenance. OMR (Optical Mark Reader) sheets are used to record the nuclear material transfer. The MUF (Materials Unaccounted For) are evaluated by PIT every three months through computer processing based on the OMR sheets. The MUF ratio of Pu handled in the facility every year from 1966 to 1977 are presented by a curve, indicating that the MUF ratio was kept well under 0.5% for every project (JOYO, FUGEN, and DCA). As for the Pu safeguards, the MBA (Material Balance Area) and the KMP (Key Measurement Point) in the facility of PNC are illustrated. The general idea of the projected PINC (Plutonium Inventory Control) system in PNC is also shortly explained. (Aoki, K.)

  4. Nuclear reactor technology progress report, vol. 4

    1981-01-01

    The works of the Engineering Section, Fast Experimental Reactor Division, are roughly classified into the technologies concerning the reactor core, abnormality monitoring, the plant, purity control and operation planning. In this paper, the activities of the Engineering Section, the operational results of Joyo and the foreign informations on FBRs in this quarter are reported. The second regular inspection carried out successively from the previous quarter was completed, and the fourth cycle operation of Joyo at 75 MW was started. The measurement of CP around the primary system pipings and equipments, the preliminary test of a core flow meter for Monju, and the various characteristic tests were carried out during this period. 2 N reports, 1 SA report and 63 memos were drawn up in this quarter. The test plan to be carried out during the period of the fourth to sixth cycle operations in this last year using the MK-1 core was formed and decided. Various meetings within and outside the division are reported. The data obtained in the operational characteristic test and special test are shown. As the results concerning the reactor technologies, the development of dosimetry techniques, the measurement and analysis of the core characteristics, the measurement of the temperature and flow velocity of coolant at the fuel assembly exit, the system pressure loss in the primary cooling system and others are reported. (Kako, I.)

  5. A review of fast reactor program in Japan. April 2000 - March 2001

    Nagata, Takashi; Yamashita, Hidetoshi

    2001-01-01

    This report describes the development and activities on fast reactors in Japan thru April 2000 to March 2001. During this period, the most important result of the Japanese Fast Reactor Project was that the first phase 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was completed at the end of March 2001, and the second phase study has just started in order to narrow down the candidate concepts selected in the first phase for next stage. In the Experimental Fast Reactor 'Joyo', the 35 th rated power operation was completed by the end of May 2000. The 13 th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) were started on the beginning of June 2000. The modification of the cooling system is underway. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage have already been drawn up based on 'Monju' comprehensive safety review. The Japan Atomic Energy Commission (JAEC) has issued a new 'Long-term Program for Research, Development and Utilization of Nuclear Energy' in November 2000. (author)

  6. The relationship of JNC and JCO in the uranium processing plant criticality accident

    Kanamori, Masashi; Yanagibashi, Katsumi; Okamoto, Naritoshi

    2002-12-01

    On September 30th 1999, the criticality accident occurred at JCO's uranium conversion building in Tokai. The accident occurred during reconversion from U 3 O 8 to uranium nitrate solution (UNH) with uranium enriched 18.8% and about 60 kgU. JCO contacted with JNC to supply UNH that is fuel material for the experimental fast breeder reactor 'JOYO'. JNC has contracted with JCO that had started nuclear fuel material processing business following a definite policy of Japanese government and developed SUMITOMO ADU PROCESS'. JNC made the first contract with JCO in 1985 and has made a contact every year. There had never been a problem in their products. JNC inspected products based on contract. JNC discharge our duty as customer inspecting products based on contract. As for safety control, JCO had taken licensing safety review and had been permitted to be 'a processing facility'. Therefore JNC understood that JCO produced following this license. 'The Uranium Processing Plant Criticality Accident Investigation' showed that JCO had been taking a different method from the permit and violating the license. However JNC had never been explained about that and JCO's operation procedures had never described about that. Therefore the Criticality Accident couldn't be avoided. This report describes the relationship of JNC and JCO in the uranium reconversion contract for JOYO, atomic development policy of Japanese government, process to the order and the contents of contract. (author)

  7. A computer analysis code of radioactive corrosion product behaviour in primary circuits of LMFBRs (PSYCHE)

    Iizawa, Katsuyuki; Seki, Seiichi; Kawasaki, Yuji; Kano, Shigeki; Nihei, Isao

    1986-01-01

    Recently it has become an important subject to reduce exposure to radiation from radioactive corrosion products (CPs) during maintenance and repair works in reactor plants. Metallic sodium is used as cooling material in fast reactor plants, leading to different CP behaviours compared to light water reactors. In the present study, a computer code for analyzing behaviours of CPs in fast reactor plants is developed. The analysis code, called PSYCHE, makes it possible to perform consistent analysis of production, migration and deposition of CPs in primary circuits together with dose rate around piping of apparatus in cooling systems. An analysis model is developed based on test results on CP behaviour in out-pile sodium. The model, called the ''dissolution-deposition model'', can reproduce atom-selective behaviour, transient phenomenon and downstream effect of CPs, which represent mass transfer phenomena in sodium. Verification of this code is carried out on the basis of CP measurements made in ''Joyo''. The calculation vs. measurement ratio is found to be 0.5 - 2 for CP deposition density in piping for cooling systems and 0.7 - 1.3 for dose rate, demonstrating that this code can give reasonable results. Analysis is also made to predict future changes in total amount of deposited CP in ''Joyo''. (Nogami, K.)

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

    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)

  9. Progress report on fast breeder reactor development in Japan

    1980-01-01

    The performance test on the reactor power increase to 75 MW was started on July 3, and the target of 75 MW was reached on July 16, in the experimental fast reactor Joyo. The tests on the heat transport characteristics, power coefficient, the response to the change of outlet temperature, the loss of external power supply and so on were carried out, and the performance test was finished on August 23, except the test of 75 MW continuous operation. The annual inspection of the systems is being carried out in parallel with the regular inspection. The design to prepare for the manufacture of the prototype fast reactor Monju is being prepared. The analysis of decay heat removal is being carried out, and various calculation codes were developed. The technological survey on overseas LMFBRs is being made. The conceptual design of the demonstration reactor is being prepared. The research and development of reactor physics, structural components for Joyo and Monju, instrumentation and control, sodium technology, fuel materials, structural materials, safety problems and steam generators are reported. The tests on the transient boiling of sodium, fuel failure propagation, heat transfer between molten materials, post-accident decay heat removal and so on have been carried out. (Kako, I.)

  10. A Review of the Fast Reactor Programme in Japan

    Matsuno, Y.

    1987-01-01

    This report covers the activities of the experimental fast reactor JOYO from April 1985 to December 1986. Five duty cycles operation from the 8th to 12th was carried out. After the 12th duty cycle operation natural circulation test from full power level of 100 MWt. The 6th annual inspection was started from the end of December 1986. Works for maintenance or modification of the plant will be done on such system as the Cesium trap in order to enable trap of Cesium which will be released from failed fuel during reactor operation, exchange of cold trap of lry coolant system in order to get a higher trapping efficiency and reloading of the on-line irradiation test assembly (INTA). Besides the operation and maintenance works, following items were conducted during the period. 1. Development of operation and maintenance supporting systems using computers. 2. Data supply to Centralized Reliability Data Organization (CREDO) of U.S.A. based on the agreement between DOE and PNC. 3. Obtaining of the license to extend the fuel maximum burnup from 50.000 MWd/t to 75.000 MWd/t. and for increase of uranium enrichment of the fuel from. 12wt% to 18wt%. Operation history of JOYO is illustrated

  11. Development of expert system on personal computer for diagnosis of nuclear reactor malfunctions

    Kameyama, Takanori; Uekata, Tomomichi; Oka, Yoshiaki; Kondo, Shunsuke; Togo, Yasumasa

    1988-01-01

    An expert system on a personal computer has been developed for diagnosis of malfunction of the fast experimental reactor 'JOYO'. Prolog-KABA is used as the language. The system diagnoses the event which causes scram or set-back of the control rod after an alarm at steady state operation. The knowledge base (KB) consists of several sub-KBs and a meta-KB. Using the forward chaining, the meta-KB decides which sub-KB should be accessed. The cause of the malfunction is identified in the sub-KB using the backward chaining. The terms expressing the characteristics of the events are involved in the production rules as attributes in order to use the Prolog function of pattern matching and back-tracking for efficient inference. The total number of the rules in the system is about 400. The experiments using the plant simulator of 'JOYO' have shown that malfunctions are successfully identified by the diagnosis system. It takes about 10s for each diagnosis using the 16-bits personal computer, PC-9801 VM. (author)

  12. Advanced analysis technology for MOX fuel

    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

  13. Global nuclear material monitoring with NDA and C/S data through integrated facility monitoring

    Howell, J.A.; Menlove, H.O.; Argo, P.; Goulding, C.; Klosterbuer, S.; Halbig, J.

    1996-01-01

    This paper focuses on a flexible, integrated demonstration of a monitoring approach for nuclear material monitoring. This includes aspects of item signature identification, perimeter portal monitoring, advanced data analysis, and communication as a part of an unattended continuous monitoring system in an operating nuclear facility. Advanced analysis is applied to the integrated nondestructive assay and containment and surveillance data that are synchronized in time. End result will be the foundation for a cost-effective monitoring system that could provide the necessary transparency even in areas that are denied to foreign nationals of both US and Russia should these processes and materials come under full-scope safeguards or bilateral agreements. Monitoring systems of this kind have the potential to provide additional benefits including improved nuclear facility security and safeguards and lower personnel radiation exposures. Demonstration facilities in this paper include VTRAP-prototype, Los Alamos Critical Assemblies Facility, Kazakhstan BM-350 Reactor monitor, DUPIC radiation monitoring, and JOYO and MONJU radiation monitoring

  14. Overview of the fast reactors fuels program

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  15. Accidents and failures in reactor facilities for test and research and reactor facilities in the stage of research and development in fiscal year 1987

    1988-01-01

    The number of accidents and failures reported in fiscal year 1987 in conformity with the law on the regulation of nuclear reactors and others was three. One case occurred during operation, and two cases occurred in shutdown state. One case was caused by improper construction management, and two cases were due to improper maintenance management. The effect of radioactivity to the surrounding environment of reactor facilities due to these accidents and failures did not arise. These occurred in the NSRR of Japan Atomic Energy Research Institute (Tokai), the experimental FBR Joyo and the ATR Fugen Power Station of Power Reactor and Nuclear Fuel Development Corp. In addition to these, the light troubles reported on the basis of the notice from the director of Science and Technology Agency dated September 1, 1981, were three cases. (K.I.)

  16. Experience of secondary cooling system modification at fast breeder reactor MONJU

    Ito, Keisuke; Nakatsuji, M.; Matsuno, Hiroki; Matsui, K.; Tone, T.

    2007-01-01

    The prototype fast breeder reactor MONJU has been shut down since the secondary sodium leak accident which occurred in December 1995. After the accident, the investigation of its cause and the comprehensive review were performed and the various counter measures against the sodium leak were also discussed. The main modification works of MONJU started in September 2005. The work should adopt suitable methods to treat sodium, since MONJU uses chemically active sodium as a coolant. Considering the chemical activity of sodium, MONJU learned the modification methods from the experimental fast reactor JOYO and precedent plants of overseas and adopted plastic bags when the sodium boundary is opened, management of oxygen concentration in the plastic bags, a slightly positive control of the cover gas pressure, compress cut by the roll cutters to prevent the entry of the chips, etc.. Owing to introduction of these methods, the modification works have proceeded almost on schedule without troubles. (author)

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

    Kisohara, Naoyuki; Sakamoto, Yoshihiko

    2010-09-01

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

  18. Overview of the fast reactors fuels program. [LMFBR

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  19. Solution treatment of fast reactor claddings

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro

    1974-01-01

    The fuel cladding tubes for Joyo (experimental FBR) are required to be a material corresponding to AISI Type 316 and cold-rolled after solution treatment. It is necessary to have no precipitation of carbide and to make the grain size smaller than ASTM No.6. It is very difficult to obtain the fine grains without the precipitation, however. In this connection, the behavior of carbide solution at high temperature and the annealing behavior of the material cold-worked and solution-treated were studied. The following matters are described: the solid solubility line of AISI Type 316; the behavior of carbide at solution treatment temperature; and the annealing behavior of the cold-worked material. (Mori, K.)

  20. Development of enclosure technique of tag gas for in-pile creep test

    Izaki, Toru; Ichikawa, Shoichi; Soroi, Masatoshi; Ito, Chikara

    2004-01-01

    Outline of the enclosure technique of tag gas for in-pile creep test is stated. In order to carry out in-pile creep test, the sample can enclose tag gas before the test and then the sample is inserted into MARICO-2 (Material Testing Rig with Temperature Control) in FBR 'JOYO' MK-III for the irradiation test. Outline of in-pile creep test using tag gas, enclosure system of tag gas, detection of a part of broken sample and identification of sample are explained. 126-, 128-, 129-, 131-, 132-, and 134-Xe are used as tag gases. The samples are identified by RIMS (Laser Resonance Ionization Mass Spectroscopy) in ppt order. ODS ferritic steel will be tested by the method in the next step. (S.Y.)

  1. Determination of irradiation temperature using SiC temperature monitors

    Maruyama, Tadashi; Onose, Shoji

    1999-01-01

    This paper describes a method for detecting the change in length of SiC temperature monitors and a discussion is made on the relationship between irradiation temperature and the recovery in length of SiC temperature monitors. The SiC specimens were irradiated in the experimental fast reactor JOYO' at the irradiation temperatures around 417 to 645degC (design temperature). The change in length of irradiated specimens was detected using a dilatometer with SiO 2 glass push rod in an infrared image furnace. The temperature at which recovery in macroscopic length begins was obtained from the annealing intersection temperature. The results of measurements indicated that a difference between annealing intersection temperature and the design temperature sometimes reached well over ±100degC. A calibration method to obtain accurate irradiation temperature was presented and compared with the design temperature. (author)

  2. Radiation shielding for fission reactors

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  3. Development of a faulty reactivity detection system applying a digital H∞ estimator

    Suzuki, Katsuo; Suzudo, Tomoaki; Nabeshima, Kunihiko

    2004-01-01

    This paper concerns an application of digital optimal H ∞ estimator to the detection of faulty reactivity in real-time. The detection system, fundamentally based on the reactivity balance method, is composed of three modules, i.e. the net reactivity estimator, the feedback reactivity estimator and the reactivity balance circuit. H ∞ optimal filters are used for these two reactivity estimators, and the nonlinear neutronics are taken into consideration especially for the design of the net reactivity estimator. A series of performance test of the detection system are conducted by using numerical simulations of reactor dynamics with the insertion of a faulty reactivity for an experimental fast breeder reactor JOYO. The system detects the typical artificial reactivity insertions during a few seconds with no stationary offset and the accuracy of 0.1 cent, and is satisfactory for its practical use. (author)

  4. Nuclear engineering experiments at experimental facilities of JNC in graduate course of Tokyo Institute of Technology

    Hayashizaki, Noriyosu; Takahashi, Minoru; Aoyama, Takafumi; Onose, Shoji

    2005-01-01

    Nuclear engineering experiments using outside facilities of the campus have been offered for graduate students in the nuclear engineering course in Tokyo Institute of Technology (Tokyo Tech.). The experiments are managed with the collaboration of Japan Nuclear Cycle Development Institute (JNC), Japan Atomic Energy Research Institute (JAERI) and Research Reactor Institute, Kyoto University (KUR). This report presents the new curriculum of the nuclear engineering experiments at JNC since 2002. The change is due to the shutdown of Deuterium Criticality Assembly Facility (DCA) that was used as an experimental facility until 2001. Reactor physics experiment using the training simulator of the experimental fast reactor JOYO is continued from the previous curriculum with the addition of the criticality approach experiment and control rods calibration. A new experimental subject is an irradiated material experiment at the Material Monitoring Facility (MMF). As a result, both are acceptable as the student experiments on the fast reactor. (author)

  5. Current trend of atomic energy development in Japan- I

    Cho, Manne; Yang, M. H.; Yoon, S. W.; Choi, M. J.; Kim, S. M.; Choi, Y. M

    1998-01-01

    After hundreds of meeting including the round table meeting on the atomic energy development policy, their conclusions were as follows: 1. The competitive nuclear fuel cycle should be completed. 2. In order to achieve above objective, the development of fast breeder reactor must be continued, and the utilization of JOYO and MONJU, and the international cooperation are highly recommended. 3. The PNC should focus on the development of the fast breeder reactor and the related fuel cycle and the management of high level radioactive waste. PNC, which had been working on too many projects, must be reformed to a slim and more efficient organization. Although there was a regret that the proposals were prepared in a short time to meet the due date for the budget bill of 1988, the Japanese government will seriously consider their proposal and take several administrative measures such as the revision of the related laws for the realization of their proposals. (author).

  6. A review of fast reactor program in Japan

    1996-01-01

    The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

  7. Studies on the dissolution of mixed oxide spent fuel from FBR

    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)

  8. Safeguards challenges of Fast Breeder Reactor

    Ko, H. S.

    2010-01-01

    Although the safeguards system of Sodium Fast Reactor (SFR) seems similar to that of Light Water Reactor (LWR), it was raised safeguards challenges of SFR that resulted from the visual opacity of liquid sodium, chemical reactivity of sodium and other characteristics of fast reactor. As it is the basic concept stage of the safeguards of SFR in Korea, this study tried to analyze the latest similar study of safeguards issues of the Fast Breeder Reactor (FBR) at Joyo and Monju in Japan. For this reason, this study is to introduce some potential safeguards challenges of Fast Breeder Reactor. With this analysis, future study could be to address the safeguards challenges of SFR in Korea

  9. Experimental research subject and renovation of chemical processing facility (CPF) for advanced fast reactor fuel reprocessing technology development

    Koyama, Tomozo; Shinozaki, Tadahiro; Nomura, Kazunori; Koma, Yoshikazu; Miyachi, Shigehiko; Ichige, Yoshiaki; Kobayashi, Tsuguyuki; Nemoto, Shin-ichi

    2002-01-01

    In order to enhance economical efficiency, environmental impact and nuclear nonproliferation resistance, the Advanced Reprocessing Technology, such as simplification and optimization of process, and applicability evaluation of the innovative technology that was not adopted up to now, has been developed for the reprocessing of the irradiated fuel taken out from a fast reactor. Renovation of the hot cell interior equipments, establishment and updating of glove boxes, installation of various analytical equipments, etc. in the Chemical Processing Facility (CPF) was done to utilize the CPF more positivity which is the center of the experimental field, where actual fuel can be used, for research and development towards establishment of the Advanced Reprocessing Technology development. The hot trials using the irradiated fuel pins of the experimental fast reactor 'JOYO' for studies on improved aqueous reprocessing technology, MA separation technology, dry process technology, etc. are scheduled to be carried out with these new equipments. (author)

  10. Doppler effect measurement in FCA assemblies X-3 and XI-1

    Okajima, Shigeaki; Mukaiyama, Takehiko

    1984-05-01

    Doppler reactivity worths were measured in FCA assemblies X-3 (mock-up core for JOYO Mark II) and XI-1 (mock-up core for large scale LMFBR) for U-238 and stractural materials of core (iron, stainless steel and nickel). The sample oscillation technique was used to measure the Doppler effect when a sample is heated up to 800 0 C from room temperature. The analysis was made using the 70 group JFS-3-J2 data set, and compared with the measured results. For U-238 samples, the calculation underestimates Doppler effects by 10%, on the other hand for other samples, the agreement between calculated values and measured values is quite good. (author)

  11. Biaxial Creep Specimen Fabrication

    JL Bump; RF Luther

    2006-01-01

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments

  12. Tokai works semi-annual progress report, July--December 1975

    1976-09-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) is a semi-governmental organization responsible for the development of advanced power reactors and nuclear fuels in Japan. The Tokai Works is the PNC center for research and development of nuclear fuels concerned with plutonium fuels fabrication, fuel reprocessing, and centrifugal uranium enrichment. Accomplishments in the activities of Tokai Works during the latter half of 1975 are summarized as follows: (1) Plutonium fuels development--Fabrication of core fuel assemblies is being continued for initial loading of the Experimental Fast Breeder Reactor JOYO and remodeling is being carried out on the facility for fabrication of plutonium fuels for the Prototype Heavy Water Moderated and Boiling Light Water Cooled Reactor FUGEN; (2) Fuel reprocessing--Construction of the Tokai Reprocessing Plant is nearly completed and preparation for its commissioning is being made; (3) Development of centrifugal uranium enrichment is being performed successfully

  13. Characterization of the insoluble sludge from the dissolution of irradiated fast breeder reactor fuel

    Aihara, Haruka; Arai, Yoichi; Shibata, Atsuhiro; Nomura, K.; Takeuchi, M. [Japan Atomic Energy Agency - JAEA, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

    2016-07-01

    Insoluble sludge is generated in the reprocessing of spent fuel. The sludge obtained from the dissolution of irradiated fuel from the Joyo experimental fast reactor was analyzed to evaluate its chemical form. The sludge was collected by the filtration of the dissolved fuel solution, and then washed in nitric acid. The yields of the sludge weight were less than 1% of the total fuel weight. The chemical composition of the sludge was analyzed after decomposition by alkaline fusion. Molybdenum, technetium, ruthenium, rhodium, and palladium were found to be the main constituent elements of the sludge. X-ray diffraction patterns of the sludge were attributable to Mo{sub 4}Ru{sub 4}RhPd, regardless of the experimental conditions. The concentrations of molybdenum and zirconium in the dissolved fast reactor fuel solutions were low, indicating that zirconium molybdate hydrate (ZMH) is produced in negligible amounts in the process. (authors)

  14. Biaxial Creep Specimen Fabrication

    JL Bump; RF Luther

    2006-02-09

    This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Naval Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.

  15. Research and development toward Monju

    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.)

  16. Heat transfer in intermediate heat exchanger under low flow rate conditions

    Mochizuki, H.

    2008-01-01

    The present paper describes the heat transfer in intermediate heat exchangers (IHXs) of liquid metal cooled fast reactors when flow rate is low such as a natural circulation condition. Although empirical correlations of heat transfer coefficients for IHX were derived using test data at the fast reactor 'Monju' and 'Joyo' and also at the 50 MW steam generator facility, the heat transfer coefficient was very low compared to the well known correlation for liquid metals proposed by Seban-Shimazaki. The heat conduction in IHX was discussed as a possible cause of the low Nusselt number. As a result, the heat conduction is not significant under the natural circulation condition, and the heat conduction term in the energy equation can be neglected in the one-dimensional plant dynamics calculation. (authors)

  17. Design considerations for economically competitive sodium cooled fast reactors

    Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

    2009-01-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

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

    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.)

  19. Fast reactor irradiation effects on fracture toughness of Si_3N_4 in comparison with MgAl_2O_4 and yttria stabilized ZrO_2

    Tada, K.; Watanabe, M.; Tachi, Y.; Kurishita, H.; Nagata, S.; Shikama, T.

    2016-01-01

    Fracture toughness of silicon nitride (Si_3N_4), magnesia-alumina spinel (MgAl_2O_4) and yttria stabilized zirconia (8 mol%Y_2O_3–ZrO_2) was evaluated by the Vickers-indentation technique after the fast reactor irradiation up to 55 dpa (displacement per atom) at about 700 °C in the Joyo. The change of the fracture toughness by the irradiation was correlated with nanostructural evolution by the irradiation, which was examined by transmission electron microscopy. The observed degradation of fracture toughness in Si_3N_4 is thought to be due to the relatively high density of small-sized of the irradiation induced defects, which should be resulted from a large amount of transmutation gases of hydrogen and helium. Observed improvement of fracture toughness in MgAl_2O_4 was due to the blocking of crack propagation by the antiphase boundaries. The radiation effects affected the fracture toughness of yttria stabilized zirconia at 55 dpa, suggesting that the generated high density voids would affect the propagation of cracks. - Highlights: • Si_3N_4, MgAl_2O_4 and YSZ were neutron irradiated up to 55dpa around 700 °C in the Joyo. • They are candidate ceramics for the inert matrices of nuclear fuels in the fast reactors. • The irradiation enhanced the fracture toughness of MgAl_2O_4 and YSZ, while degraded that of Si_3N_4. • The toughness changes were correlated with radiation induced defects and transmutation gases.

  20. Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

    Fukuda, Makoto, E-mail: makoto.fukuda@qse.tohoku.ac.jp [Tohoku University, Sendai, 980-8579 (Japan); Kiran Kumar, N.A.P.; Koyanagi, Takaaki; Garrison, Lauren M. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA, 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Hasegawa, Akira [Tohoku University, Sendai, 980-8579 (Japan)

    2016-10-15

    Neutron irradiation to single crystal pure tungsten was performed in the mixed spectrum High Flux Isotope Reactor (HFIR). To investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ∼90–∼800 °C and fast neutron fluences were 0.02–9.00 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. The hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten. - Highlights: • The microstructure and irradiation hardening of single crystal pure W irradiated in HFIR was investigated. • The neutron energy spectrum influence was evaluated by comparing the HFIR results with previous work in Joyo and JMTR. • In the dose range up to ∼1 dpa, the neutron energy spectrum influence of irradiation hardening was not clear. • In the dose range above 1 dpa, the neutron energy influence on irradiation hardening and microstructural development was clearly observed. • The irradiation induced precipitates caused significant irradiation hardening of pure W irradiated in HFIR.

  1. Development and verifications of fast reactor fuel design code ''Ceptar''

    Ozawa, T.; Nakazawa, H.; Abe, T.

    2001-01-01

    The annular fuel is very beneficial for fast reactors, because it is available for both high power and high burn-up. Concerning the irradiation behavior of the annular fuel, most of annular pellets irradiated up to high burn-up showed shrinkage of the central hole due to deformation and restructuring of the pellets. It is needed to predict precisely the shrinkage of the central hole during irradiation, because it has a great influence on power-to-melt. In this paper, outline of CEPTAR code (Calculation code to Evaluate fuel pin stability for annular fuel design) developed to meet this need is presented. In this code, the radial profile of fuel density can be computed by using the void migration model, and law of conservation of mass defines the inner diameter. For the mechanical analysis, the fuel and cladding deformation caused by the thermal expansion, swelling and creep is computed by the stress-strain analysis using the approximation of plane-strain. In addition, CEPTAR can also take into account the effect of Joint-Oxide-Gain (JOG) which is observed in fuel-cladding gap of high burn-up fuel. JOG has an effect to decrease the fuel swelling and to improve the gap conductance due to deposition of solid fission product. Based on post-irradiation data on PFR annular fuel, we developed an empirical model for JOG. For code verifications, the thermal and mechanical data obtained from various irradiation tests and post-irradiation examinations were compared with the predictions of this code. In this study, INTA (instrumented test assembly) test in JOYO, PTM (power-to-melt) test in JOYO, EBR-II, FFTF and MTR in Harwell laboratory, and post-irradiation examinations on a number of PFR fuels, were used as verification data. (author)

  2. A review of fast reactor program in Japan. April 1997 - March 1998

    1998-01-01

    This report describes the development and activities on fast reactor in Japan for the period of April 1997 - March 1998. During this period, two important results were drawn by the Special Committee on Fast Breeder Reactors (FBRs) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) Reform Committee, respectively. The Special Committee on FBRs discussed on the future FBR development in Japan including the Prototype FBR 'Monju' operation, and proposed its conclusion as the final report to the Japan Atomic Energy Commission (JAEC) on December 1, 1997. The PNC Reform Committee reviewed PNC's management and safety assurance system, and recommended to reform PNC to a new organization. Each committee result is outlined in this report. The Experimental Fast Reactor 'Joyo' operated 30th - 32nd cycle. In parallel with the operation, the Joyo Upgrading Program (MK-III program) is in progress. Five MK-III driver fuel subassemblies were loaded to the core in the 32nd cycle. Monju comprehensive safety review, which was started in December 1996, was continued through 1997, and was completed in March 1998. The DFBR Plant Optimization (phase 2) design study was launched by the Japan Atomic Power Company (JAPC) with goal of constructing FBR plant that achieves both reliability and economy from FY 1997 for three years. Research and development works are underway under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of PNC, JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). (author)

  3. Plug-welding of ODS cladding tube for BOR-60 irradiation. Welding condition setting. Device remodeling and welding

    Seki, Masayuki; Ishibashi, Fujio; Kono, Syusaku; Hirako, Kazuhito; Tsukada, Tatsuya

    2003-04-01

    Irradiation test in BOR-60 at RIAR to judge practical use prospect of ODS cladding tube at early stage is planned as Japan-Russia a joint research. RIAR does fuel design of fuel pin used for this joint research. JNC manufactures ODS cladding tube and bar materials (two steel kind of martensite and ferrite), upper endplug production. They are welded by pressurized resistance welding, and are inspected in JNC Tokai, transported to RIAR. And RIAR manufactures vibration packing fuel pin. On the upper endplug welding by pressurized resistance welding method, we worded on the problems such as decision of welding condition by changing the size and crystallization of cladding tube and the design of endplug, and the chucking device remodeling to correspond to the long scale cladding tube welding system (included handling) and of quality assurance method. Especially, use of long scale cladding tube caused problem that bending transformation occurred in cladding tube by welding pressure. However, we solved this problem by shortening the distance of cladding tube colette chuck and pressure receiving, and by putting the sleeve in an internal space of welding machine, losing the bending of cladding tube. Moreover, welding defects were occurred by the difference of an inside state, an inside defect and recrystallization of cladding tube. We solved the problem by inside grinding for the edge of tube, angle beam method by ultrasonic wave, and ultrasonic wave form confirmation. Manufacturing process with long scale cladding tube including heat-treatment to remove combustion return and remaining stress was established besides, Afterwards, welding of ODS cladding tube and upper endplug. As the quality assurance system, we constructed [Documented procedure (referred to JOYO)] based on [Document of the QA plan] by OEC. Welding and inspection were executed by the document procedure. It is thought that the quality assurance method become references for the irradiation test in JOYO in the

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

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

    2009-01-01

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

  5. Investigation of safety measures to severe accident of Fast Breeder Reactor

    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)

  6. Property change of advanced tungsten alloys due to neutron irradiation

    Fukuda, Makoto; Hasegawa, Akira; Tanno, Takashi; Nogami, Shuhei; Kurishita, Hiroaki

    2013-01-01

    This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073 K and 0.15–0.47 dpa, respectively. TEM images of all samples after 0.42 dpa irradiation at 1023 K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5 wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening

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

    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.)

  8. Overall plant concept for a tank-type fast reactor

    Yamaki, Hideo; Davies, S.M.; Goodman, L.

    1984-01-01

    Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

  9. Results of Am isotopic ratio analysis in irradiated MOX fuels

    Koyama, Shin-ichi; Osaka, Masahiko; Mitsugashira, Toshiaki; Konno, Koichi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Kajitani, Mikio

    1997-04-01

    For analysis of a small quantity of americium, it is necessary to separate from curium which has similar chemical property. As a chemical separation method for americium and curium, the oxidation of americium with pentavalent bismuth and subsequent co-precipitation of trivalent curium with BIP O{sub 4} were applied to analyze americium in irradiated MOX fuels which contained about 30wt% plutonium and 0.9wt% {sup 241}Am before irradiation and were irradiated up to 26.2GWd/t in the experimental fast reactor Joyo. The purpose of this study is to measure isotopic ratio of americium and to evaluate the change of isotopic ratio with irradiation. Following results are obtained in this study. (1) The isotopic ratio of americium ({sup 241}Am, {sup 242m}Am and {sup 243}Am) can be analyzed in the MOX fuels by isolating americium. The isotopic ratio of {sup 242m}Am and {sup 243}Am increases up to 0.62at% and 0.82at% at maximum burnup, respectively, (2) The results of isotopic analysis indicates that the contents of {sup 241}Am decreases, whereas {sup 242m}Am, {sup 243}Am increase linearly with increasing burnup. (author)

  10. Philosophy of safety evaluation on fast breeder reactor

    1981-01-01

    This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

  11. Model-based temperature noise monitoring methods for LMFBR core anomaly detection

    Tamaoki, Tetsuo; Sonoda, Yukio; Sato, Masuo; Takahashi, Ryoichi.

    1994-01-01

    Temperature noise, measured by thermocouples mounted at each core fuel subassembly, is considered to be the most useful signal for detecting and locating local cooling anomalies in an LMFBR core. However, the core outlet temperature noise contains background noise due to fluctuations in the operating parameters including reactor power. It is therefore necessary to reduce this background noise for highly sensitive anomaly detection by subtracting predictable components from the measured signal. In the present study, both a physical model and an autoregressive model were applied to noise data measured in the experimental fast reactor JOYO. The results indicate that the autoregressive model has a higher precision than the physical model in background noise prediction. Based on these results, an 'autoregressive model modification method' is proposed, in which a temporary autoregressive model is generated by interpolation or extrapolation of reference models identified under a small number of different operating conditions. The generated autoregressive model has shown sufficient precision over a wide range of reactor power in applications to artificial noise data produced by an LMFBR noise simulator even when the coolant flow rate was changed to keep a constant power-to-flow ratio. (author)

  12. Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging*1

    Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

    1988-07-01

    Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1 × 10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6C in JPCA-2.

  13. Chemical surveillance of commercial fast breeder reactors

    Stamm, H.H.; Stade, K.Ch.

    1988-01-01

    After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

  14. Uranium, Plutonium and Neptunium Co-recovery with Irradiated Fast Reactor MOX Fuel by Single Cycle Extraction Process

    Masaumi Nakahara; Yuichi Sano; Kazunori Nomura; Tadahiro Washiya; Jun Komaki [Japan Atomic Energy Agency, 4-33 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1194 (Japan)

    2008-07-01

    The behavior of Np in single cycle extraction processes using tri-n-butylphosphate (TBP) as an extractant for U, Pu and Np co-recovery was investigated as a part of NEXT (New Extraction System for Transuranium) process. Two approaches for Np co-recovery with U and Pu were carried out with irradiated MOX fuel from fast reactor 'JOYO'; one was the counter current experiment using a feed solution with a high HNO{sub 3} concentration and the other used a scrubbing solution with a high HNO{sub 3} concentration. Experimental results showed that the leakage of Np to the raffinate were 0.986 % and 5.96 % under the condition of high HNO{sub 3} concentration in the feed solution and scrubbing solution, respectively. The simulation results based on these experiments indicated that most of Np could be extracted and co-recovered with U and Pu, just by increasing HNO{sub 3} concentrations in the feed and scrubbing solution on the single cycle extraction process. (authors)

  15. Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging

    Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

    1988-01-01

    Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1x10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6 C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6 C in JPCA-2. (orig.)

  16. Role of fast breeders in Japan

    Oyama, A.; Tomabechi, K.

    1978-09-01

    To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tons, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being operated at 50MW and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction with a schedule of operation in the mid-1980s. Following operation of MONJU, construction of a large demonstration reactor of 1000 - 15000 MW(e) will be undertaken. Some 2 - 3 years after the construction of the demonstration reactor, a series of reactors will be constructed similar in size and design to promote commercialization of LMFBRs. Strong efforts will be made to put this programme into practice. It is expected that LMFBRs will play an important role in mitigating the serious problem of energy supply in Japan foreseeable around the turn of the century

  17. Irradiation behavior of graphite shielding materials for FBR

    Maruyama, Tadashi; Kaito, Takeji; Onose, Shoji; Shibahara, Itaru

    1994-01-01

    Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor 'JOYO' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597degC. Postirradiation examination was carried out on dimensional change, elastic modulus, and the thermal conductivity. The result of measurement of dimensional change indicated that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased to two to three times of unirradiated values. A large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependency on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, but the change in specific heat was negligibly small. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (author)

  18. Progress report on fast breeder reactor development in Japan, October - December 1976

    1977-03-01

    As for the fast breeder experimental reactor ''Joyo'', the remodeling works for the bearings of primary main circulation pump and the support bands of hanger support have finished. Preparatory arrangement was made for the second performance test, such as preheating the primary cooling system and filling sodium in it. Construction, installation and adjustment of test and inspection apparatuses required after the criticality experiment are in progress. The lining work for the temporary storage pool for solid wastes, the enlarging work for the operation control building, and construction of the brake equipment for cooler blower were over. As for the operation control, operation, inspection and maintenance of each system required for the modeling work are being carried out, and new parts to take the place of old ones were provided. The analysis of the core characteristics and the development of operation-monitoring codes are in progress. The coordination design work (4) for the fast breeder prototype reactor ''Monju'' has been started, and the preliminary design work (2) for the demonstration reactor is going on. Geological, meteorological and seismic researches carried out at the proposed construction site for ''Monju''. The researches and developments on reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, safety and steam generators are in progress. (Kako, I.)

  19. Current status of FBR development in Japan

    Ichimiya, Masakazu

    2008-01-01

    'Fast Reactor Cycle Technology Development (FaCT)' project has been conducted since 2006. In this project, design study and research and development (R and D) on innovative technologies for fast reactor (FR) cycle system are implemented in order to present the conceptual designs of commercial and demonstration facilities by 2015 and start operating demonstration fast reactor in 2025. The R and Ds has been stepped forward into the development stage to establish the realization of innovative technologies which bring excellent performance to fast reactor cycle system. The purpose of R and D by 2010 is to decide whether innovative technologies shall be adopted. In the FaCT project, R and D stage for the realization of innovative technologies, it is important to take full advantage of JAEA's R and D facilities toward demonstration and commercialization stages. Particularly, it is indispensible for the realization of innovative technologies to develop the fuel and material by irradiation tests using an experimental reactor 'Joyo', to verify the reliability for power-generating plant through the experience of operation and maintenance, and to establish technologies of operation, maintenance and repair for the plant with a prototype reactor 'Monju'. Several possible R and D have been effectively carried out within the frameworks of international cooperation, such as Global Nuclear Energy Partnership (GNEP), Generation IV International Forum (GIF), and International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). (author)

  20. Towards the inclusion of open fabrication porosity in a fission gas release model

    Claisse, Antoine, E-mail: claisse@kth.se [KTH Royal Institute of Technology, Reactor Physics, AlbaNova University Centre, 106 91, Stockholm (Sweden); Van Uffelen, Paul [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125, Karlsruhe (Germany)

    2015-11-15

    A model is proposed for fission product release in oxide fuels that takes into account the open porosity in a mechanistic manner. Its mathematical framework, assumptions and limitations are presented. It is based on the model for open porosity in the sintering process of crystalline solids. More precisely, a grain is represented by a tetrakaidecahedron and the open porosity is represented by a continuous cylinder along the grain edges. It has been integrated in the TRANSURANUS fuel performance code and applied to the first case of the first FUMEX project as well as to neptunium and americium containing pins irradiated during the SUPERFACT experiment and in the JOYO reactor. The results for LWR and FBR fuels are consistent with the experimental data and the predictions of previous empirical models when the thermal mechanisms are the main drivers of the release, even without using a fitting parameter. They also show a different but somewhat expected behaviour when very high porosity fuels are irradiated at a very low burn-up and at low temperature. - Highlights: • We developed a new athermal FGR model based on the porosity. • We present the model, its framework, assumptions and limitations. • We test it out on several irradiation experiments. • Results are comparable to previous models but without using an empirical parameter.

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

    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.)

  2. Extraction residue analysis on F82H-BA07 heat and other reduced activation ferritic/martensitic steels

    Nagasaka, Takuya; Hishinuma, Yoshimitsu; Muroga, Takeo; Li, Yanfen; Watanabe, Hideo; Tanigawa, Hiroyasu; Sakasegawa, Hideo; Ando, Masami

    2011-01-01

    Extraction residue analysis was conducted on reduced activation ferritic/martensitic steels, such as F82H-BA07 heat, F82H-IEA heat, JLF-1 JOYO heat and CLAM steel. M 23 C 6 type precipitates, TaC precipitates and Fe 2 W Laves phase were identified in the present analyses. M 23 C 6 precipitates were coarsened in F82H-BA07 compared with the other steels at as-normalized and tempered (NT) condition. TaC precipitate formation was enhanced in JLF-1 and CLAM compared with F82H-BA07 and F82H-IEA at as-NT condition. Laves phase were detected in F82H-IEA after aging above 550 o C, where solid solution W was significantly decreased. F82H-IEA exhibited hardening after aging at 400 and 500 o C for 100 khr, whereas softening at 600 and 650 o C. This behavior is similar to JLF-1 and CLAM, and can be understood by precipitation of TaC and Laves phase.

  3. Simulating irradiation hardening in tungsten under fast neutron irradiation including Re production by transmutation

    Huang, Chen-Hsi; Gilbert, Mark R.; Marian, Jaime

    2018-02-01

    Simulations of neutron damage under fusion energy conditions must capture the effects of transmutation, both in terms of accurate chemical inventory buildup as well as the physics of the interactions between transmutation elements and irradiation defect clusters. In this work, we integrate neutronics, primary damage calculations, molecular dynamics results, Re transmutation calculations, and stochastic cluster dynamics simulations to study neutron damage in single-crystal tungsten to mimic divertor materials. To gauge the accuracy and validity of the simulations, we first study the material response under experimental conditions at the JOYO fast reactor in Japan and the High Flux Isotope Reactor at Oak Ridge National Laboratory, for which measurements of cluster densities and hardening levels up to 2 dpa exist. We then provide calculations under expected DEMO fusion conditions. Several key mechanisms involving Re atoms and defect clusters are found to govern the accumulation of irradiation damage in each case. We use established correlations to translate damage accumulation into hardening increases and compare our results to the experimental measurements. We find hardening increases in excess of 5000 MPa in all cases, which casts doubts about the integrity of W-based materials under long-term fusion exposure.

  4. An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

    Osaka, Masahiko; Koyama, Shin-ichi; Maeda, Shigetaka; Mitsugashira, Toshiaki

    2005-01-01

    Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the 241 Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the 240 Pu to 239 Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor

  5. Summary reports of doctoral research fellows in 1999

    2000-07-01

    JNC (Japan Nuclear Cycle Development Institute) has been promoting the fellowship program for young scientists to encourage their expertise from 1997. The objective of this program is to raise the talented persons for future JNC activities. This report summarized the status on the 17 themes of researches implemented by young research fellows in 1999. The report includes the list of individual titles and the site and supporting staffs of JNC. The individual reports given by 17 authors describe its time schedule, status, achievements and own publications in the prearranged format. Topics include the followings; Those are giant resonance structures of short lifetime nuclei, growth of cavities in structural materials, calculation methods of photonuclear interaction cross-section, stability and radioactivity diffusion of bentonite, plant ecology of radioactivity in uranium mines, numerical simulations on the TRUEX process, advanced nondestructive testing of materials, physical properties of actinides obtained by band theory, de-oxidisation capability in pyrite to decrease oxygen in bentonite, investigation on equilibrium fast reactor recycling, behaviors of underground water in the Tono Mine, effects of radiation induced educts on swelling, simulation on fuel-coolant interaction after severe accidents, and resonance ionization mass spectrometry for the JOYO reactor to detect failed fuel locations. (Tanaka, Y.)

  6. Joint research achievement report on field test project for photovoltaic power generation of industrial use in fiscal 2000; 1999 nendo sangyo tou you taiyokohatsuden field test jigyo kyodo kenkyu seika hokokusho

    NONE

    2001-05-01

    This paper summarizes the achievements in fiscal 2000 on the field tests for photovoltaic power generation of industrial use. This report describes the details of the achievements of the field tests in 36 locations including the following organizations: the Japan Industrial University, a school juristic person; the Harukawa Gakusha, a school juristic person; the Snow Brand Dairy; the Sony Chemicals; the West Japan Passenger Railways; the Hokushin Welfare Society, a social welfare juristic person; the Seikyou Gakuen, a school juristic person; the Miko Corporation; the Akashi Apparel Corporation; Shigenobu Township in Ehime Prefecture; the Japan Ham; the Seiko Epson; Henri Charpantier Corporation; the Joyo Gakuen, a school juristic person; Matsushita Electric Industry; Nio Township in Kagawa Prefecture; Matsumae Township in Ehime Prefecture; Nakajima Township in Ehime Prefecture; Sugo Gakuen, a school juristic person; the East Japan East Japan Passenger Railways (General Training Center); Shizuoka Automobile Gakuen, a school juristic person; Tomiyama Village in Aichi Prefecture; Futamura Grinding Industry; Toba City in Mie Prefecture; Higashi-ura Township in Hyogo Prefecture, Taisho Township in Kochi Prefecture; the Nagasaki Branch Office of the West Japan Telegraph and Telephone Company; Ashikita Township in Kumamoto Prefecture; the Miyazaki Branch Office of the West Japan Telegraph and Telephone Company; Oi Township in Fukui Prefecture; the Keishin Society, a social welfare juristic person; the Azabu Veterinarian School, a school juristic person, Omiya Township in Ibaraki Prefecture; and the Showa Electric Machinery. (NEDO)

  7. A review of fast reactor programme in Japan

    1998-01-01

    This report describes the development and activities on fast reactor in Japan for the period of April 1996 - March 1997. During this period, the 30th duty cycle operation has been started in the Experimental Fast Reactor ''''Joyo''''. The cause investigation on the sodium leak incident has completed and the safety examination are being performed in the Prototype Fast Breeder Reactor ''''Monju''''. The three years design study since FY1994 on the plant optimization of the Demonstration FBR has been completed by the Japan Atomic Power Company (JAPC). Related research and development works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which is composed of Power Reactor and Nuclear Fuel Development Corporation (PNC), JAPC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). In November 1996, the Japan Atomic Energy Commission (JAEC) established a Social Gathering Meeting to discuss generally the significance of FBR development in Japan for the future. (author)

  8. Applicability study of optical fiber distribution sensing to nuclear facilities

    Takada, Eiji; Kimura, Atsushi; Nakazawa, Masaharu; Kakuta, Tsunemi

    1999-01-01

    Optical fibers have advantages like flexible configuration, intrinsic immunity for electromagnetic fields etc., and they have been used for signal transmission and as optical fiber sensors (OFSs). By some of these sensor techniques, continuous or discrete distribution of physical parameters can be measured. Here, in order to discuss the applicability of these OFSs to nuclear facilities, irradiation experiments to optical fibers were carried out using the fast neutron source reactor 'YAYOI' and a 60 Co γ source. It has been shown that, under irradiation with fast neutrons, the radiation induced loss increase almost linearly with the neutron fluence. On the other hand, when irradiated with 60 Co γ rays, the loss shows a saturation tendency. As an example of the OFSs, applicability of the Raman distributed temperature sensor (RDTS) to the monitoring of nuclear facilities has been examined. Two correction techniques for radiation induced errors have been developed and for the demonstration of their feasibility, measurements were carried out along the primary piping system of the experimental fast reactor: JOYO. During the continuous measurements with the total dose of more than 10 7 [R], the radiation induced errors showed a saturating tendency and the feasibility of the loss correction technique was demonstrated. Although the time response of the system should be improved, the RDTS can be expected as a noble temperature monitor in nuclear facilities. (author)

  9. Development of a fine and ultra-fine group cell calculation code SLAROM-UF for fast reactor analyses

    Hazama, Taira; Chiba, Go; Sugino, Kazuteru

    2006-01-01

    A cell calculation code SLAROM-UF has been developed for fast reactor analyses to produce effective cross sections with high accuracy in practical computing time, taking full advantage of fine and ultra-fine group calculation schemes. The fine group calculation covers the whole energy range in a maximum of 900-group structure. The structure is finer above 52.5 keV with a minimum lethargy width of 0.008. The ultra-fine group calculation solves the slowing down equation below 52.5 keV to treat resonance structures directly and precisely including resonance interference effects. Effective cross sections obtained in the two calculations are combined to produce effective cross sections over the entire energy range. Calculation accuracy and improvements from conventional 70-group cell calculation results were investigated through comparisons with reference values obtained with continuous energy Monte Carlo calculations. It was confirmed that SLAROM-UF reduces the difference in k-infinity from 0.15 to 0.01% for a JOYO MK-I fuel subassembly lattice cell calculation, and from - 0.21% to less than a statistical uncertainty of the reference calculation of 0.03% for a ZPPR-10A core criticality calculation. (author)

  10. An examination of reliability critical items in liquid metal reactors: An analysis by the Centralized Reliability Data Organization (CREDO)

    Humphrys, B.L.; Haire, M.J.; Koger, K.H.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.; Okubo, Y.

    1987-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 21,000 components and approximately 1300 event records. A conservative estimation is that the total component operating hours is approaching 3.5 billion hours. Because data gathering for CREDO concentrates on event (failure) information, the work reported here focuses on the reliability information contained in CREDO and the development of reliability critical items lists. That is, components are ranked in prioritized lists from worst to best performers from a reliability standpoint. For the data contained in the CREDO data base, FFTF and JOYO show reliability growth; EBR-II reveals a slight unreliability growth for those components tracked by CREDO. However, tabulations of events which cause reactor shutdowns decrease with time at each site

  11. Centralized Reliability Data Organization (CREDO) assessment of critical component unavailability in liquid metal reactors

    Koger, K.H.; Haire, M.J.; Humphrys, B.L.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.

    1988-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Dept. of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 20,000 components and approximately 1500 event records. A conservative estimation is that the total component operating hours is approaching 2.2 billion hours. The work reported here focuses on the availability information contained in CREDO and the development of availability critical items lists. That is, individual components are ranked in prioritized lists from worst to best performers from an availability standpoint. Availability as used here is an inherent characteristics of the component and is not necessarily related to plant operability. A major observation is that a few components have a much higher unavailability factor than the average. The top fifteen components contribute 93%, 77%, and 87% of the total system unavailability for EBR-II, FFTF, and JOYO respectively. Critical components common to all three sites are mechanical pumps and electromagnetic pumps. Application of resources to these components with the highest unavailability will have the greatest effect on overall availability. All three sites demonstrate that low maintainability (i.e., long repair times), rather than unreliability (i.e., high failure rates), are the main contributors, by about a two-to-one margin, to liquid metal system unavailability

  12. The instrumentation of fast reactor

    Endo, Akira

    2003-03-01

    The author has been engaged in the development of fast reactors over the last 30 years with both an involvement with the early technology development on the experimental breeder reactor Joyo, and latterly continuing this work on the prototype breeder reactor, Monju. In order to pass on this experience to younger engineers this paper is produced to outline this experience in the sincere hope that the information given will be utilised in future educational training material. The paper discusses the wide diversity on the associated instrument technology which the fast breeder reactor requires. The first chapter outlines the fast reactor system, followed by discussions on reactor instrumentation, measurement principles, temperature dependencies, and verification response characteristics from various viewpoints, are discussed in chapters two and three. The important issues of failed fuel location detection, and sodium leak detection from steam generators are discussed in chapters 4 and 5 respectively. Appended to this report is an explanation on the methods of measuring response characteristics on instrumentation systems using error analysis, random signal theory and measuring method of response characteristic by AR (autoregressive) model on which it appears is becoming an indispensable problem for persons involved with this technology in the future. (author)

  13. Past and present role of fast breeder reactors in Japan

    Tomabechi, K.

    1978-01-01

    To meet increasing future energy demand in Japan utilization of fission energy should be promoted. In particular it is of vital importance to develop and utilize FBRs as soon as possible in order to save the natural uranium needed. If one considers the commercial introduction of FBRs in the mid-1990s in Japan, a delay of only one year will eventually result in an additional demand for natural uranium of more than 20,000 tonnes, because several LWRs will have to be installed instead. Ten years have passed since the development of FBRs in Japan was initiated as a national project with the highest priority and now the experimental fast reactor JOYO is successfully being commissioned and the prototype fast breeder reactor MONJU has reached the stage of proceeding to construction. Experience gained from past development work encourages further strong effort towards the commercialization of FBRs. This paper briefly reviews the important role to be played by FBRs, the development programme for FBRs, and the experience gained so far from the programme in Japan. (author)

  14. Development of a multi-functional reprocessing process based on ion-exchange method by using tertiary pyridine-type resin

    Koyama, Shin-ichi; Ozawa, Masaki; Suzuki, Tatsuya; Fujii, Yasuhiko

    2006-01-01

    A series of separation experiment was performed in order to study a multi-functional spent fuel reprocessing process based on ion-exchange technique. The tertiary pyridine-type anion-exchange resin was used in this experiment and the mixed oxide fuel highly irradiated in the experimental fast reactor ''JOYO'' was used as a reference spent fuel. As the result, 106 Ru + 125 Sb, 137 Cs + 155 Eu + 144 Ce, plutonium, americium and curium could be separated from the irradiated fuel by only three steps of ion-exchange. The decontamination factor of 137 Cs and trivalent lanthanides ( 155 Eu, 144 Ce) in the final americium product exceeded 3.9 x 10 4 and 1.0 x 10 5 , respectively. The decontamination factor for the mutual separation of 243 Cm and 241 Am was larger than 2.2 x 10 3 for the americium product and, moreover, the content of 137 Cs, trivalent lanthanides and 243 Cm included in 241 Am product did not exceed 2 ppm. These results prove that the proposed simplified separation process has a reality as a candidate for future reprocessing process based on the partitioning and transmutation concept. (author)

  15. A Review of the Fast Reactor Programme in Japan

    Matsuno, Y.

    1988-01-01

    This report covers the activities of the experimental fast reactor JOYO from April 1986 to December 1987. Five duty cycles operations from the 10th to 14th were done. Starting from the 13th operation, the duration per cycle was extended to 55 days at 13th cycle and 60 days at 14th, in order to elongate the fuel maximum burn-up from 50,000 MWd/t to 75,000 MWd/t. By the end of 15th cycle (mid-May of '88), the objective 70 days duration is attained. The 6th annual inspection was finished in September 1987. Work for maintenance or modification of the plant has been done on such system as Cecium trap in order to enable the loop to trap the Cecium which will be released from failed fuel during the reactor operation, the cold trap (replacement) of the primary coolant system in order to get higher trapping efficiency and the 2nd set-up (reloading) of on-line irradiation test assembly (INTA-S)

  16. Progress report on fast breeder reactor development in Japan

    1979-01-01

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

  17. Progress report on fast breeder reactor development in Japan, July - September 1977

    1978-06-01

    As for the experimental fast breeder reactor ''Joyo'', the low power performance tests have been continued, and the measurements of reactor noise, the reactivity of fuel assemblies, power distribution and Na-void effect have been made. Efforts have been exerted to develop the required maintenance equipments, to manufacture the transfer rotor maintenance facilities, and to construct the spent fuel storing and cooling facilities. The analysis and calculation of the core characteristics have been in progress. The design work on the prototype fast breeder reactor ''Monju'' has been continued, and the development of the computer codes for the design has progressed. Informations have been gathered regarding the technological developments of LMFBRs overseas. The surveys on the site for ''Monju'' have been carried out. The design and research works on the demonstration reactor were started, and the general design factors such as the steam condition and the plant layout have been studied. As for the research and development of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety, and steam generators, the progresses are reported in detail. High performance neutron detectors for nuclear instrumentation have been under development, and the tagging gas method for fuel failure detection and location system has been tested. (Kako, I.)

  18. Progress report on fast breeder reactor development in Japan, June-March, 1979

    1979-11-01

    The experimental fast reactor ''Joyo'' started the second cycle operation of 50 MW on January 12, and finished it on February 26. The operation was very stable throughout this period. The preparation to raise the output to 75 MW has been in progress since then in parallel with the periodic inspection. Experimental fuel subassemblies and a control rod used for the 50 MW operation were removed. Installations of a fuel storage and handling facility, a water cooling and purifying plant, and an in-water cutting equipment were completed. The works related to the analysis of the core characteristics are reported. The construction preliminary design (2) of the prototype fast reactor ''Monju'' was finished. The wind tunnel experiment and the calculation of earthquake response to artificial seismic waves are being carried on. The works of developing codes and the surveys on the construction site are reported. The fourth preliminary design of the demonstration fast reactor was completed. The research and development of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and structural materials, safety and steam generators are reported. The technological investigation into foreign LMFBRs was finished. (Kako, I.)

  19. Irradiation Assisted Stress Corrosion Cracking of austenitic stainless steels

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) of austenitic stainless steels in oxygenated high temperature water was studied. The IASCC failure has been considered as a degradation phenomenon potential not only in the present light water reactors but rather common in systems where the materials are exposed simultaneously to radiation and water environments. In this study, effects of the material and environmental factors on the IASCC of austenitic stainless steels were investigated in order to understand the underlying mechanism. The following three types of materials were examined: a series of model alloys irradiated at normal water-cooled research reactors (JRR-3M and JMTR), the material irradiated at a spectrally tailored mixed-spectrum research reactor (ORR), and the material sampled from a duct tube of a fuel assembly used in the experimental LMFBR (JOYO). Post-irradiation stress corrosion cracking tests in a high-temperature water, electrochemical corrosion tests, etc., were performed at hot laboratories. Based on the results obtained, analyses were made on the effects of alloying/impurity elements, irradiation/testing temperatures and material processing, (i.e., post-irradiation annealing and cold working) on the cracking behavior. On the basis of the analyses, possible remedies against IASCC in the core internals were discussed from viewpoints of complex combined effects among materials, environment and processing factors. (author). 156 refs.

  20. Development of radioactivity estimation system considering radioactive nuclide movement

    Fukumura, Nobuo; Miyamoto, Yoshiaki

    2010-01-01

    A radioactivity estimation system considering radioactive nuclide movement is developed to integrate the established codes and the code system for decommissioning of sodium cooled fast reactor (FBR). The former are the codes for estimation of radioactivity movement in sodium coolant of fast reactor which are named SAFFIRE, PSYCHE and TTT. The latter code system is to estimate neutron irradiation activity (COSMARD-RRADO). It is paid special attention to keep the consistency of input data used among these codes and also the simplification of their interface. A new function is added to the estimation system, to estimate minor FP inventory caused by the fission of impurities contained in the coolant and slight fuel material attached on the fuel cladding. To check the evaluation system, the system is applied with radioactivity data of the preceding FBR such as BN-350, JOYO and Monju. Agreement between the analysis results and the measurement is well satisfactory. The uncertainty of the code system is within several tens per cent for the activation of primary coolant (Na-22) and factor of 2-4 for the estimation of radioactivity inventory in sodium coolant. (author)

  1. Proceedings of 2005 JAEA-KAERI joint seminar on advanced irradiation and PIE technologies

    2006-05-01

    In this seminar, total participants of over 100 were jointed from JAEA, KAERI, Hanyang University, Chungnam National University, Kyung Hee University, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. The technical development and experimental data on the irradiation test and PIE were aggressively discussed in this seminar. Contributed presentations were 35 in three sessions; Current status and future program on irradiation test and PIE (10 presentations), Development of irradiation and PIE technologies (15 presentations) and Evaluation of irradiation and PIE data (10 presentations). Development of instrumented capsule technologies for HANARO irradiation, current PIE activities in each hot laboratory of both countries, development of irradiation capsules in JMTR for the Irradiation Assisted Stress Corrosion Cracking (IASCC) study, development of irradiation and PIE techniques for the safety research on the high burnup fuel, utilization plan of JOYO and development of MOX fuel containing americium have been widely noticed as topic items on irradiation and PIE technologies. This proceedings is containing papers presented in the 2005 JAEA-KAERI Joint Seminar. It also indicates the current status of the aggressive information exchange activity on two fields of irradiation test and PIE technologies between JAEA and KAERI under the Arrangement for the Implementation of Cooperative Research Program mentioned above. The 35 of the presented papers are indexed individually. (J.P.N.)

  2. Yearly program of safety research for nuclear facilities and others

    1987-01-01

    The development of FBRs in Japan has steadily progressed, and subsequently to the experimental reactor 'Joyo' and the prototype reactor 'Monju', by promoting the construction of a demonstration reactor, the stage of verifying and acquiring skill of the electricity generation plant technology of practical scale, improving the performance and establishing the economical efficiency is about to begin. The development of FBRs in Japan has been advanced independently as a national project, and the method of preventing accidents in the actual reactors has been thoroughly taken. 'On the way of thinking in the safety evaluation of FBRs' was decided by the Nuclear Safety Commission. When the safety research from 1987 is systematized, as the constituents of safety logic, the way of thinking of the defense in depth, the way of thinking of the classification according to importance, the way of thinking of multilayer barriers against radioactive substances, and the way of thinking on severe accidents were investigated. The research concerning the decision of safety design and evaluation policy, and the safety research regarding accident prevention and relaxation, accident evaluation and severe accidents are reported. (Kako, I.)

  3. Development on the ex-vessel transfer machine for Monju, (1)

    Inoue, Takashi; Kinuta, Kenji; Tomita, Takaaki

    1980-01-01

    As for the fast breeder reactors being developed as the national project, the construction of the prototype reactor ''Monju'' is started in 1980 based on the results of development of the experimental reactor ''Joyo''. In the fuel-handling facility, the development of which is promoted by Fuji Electric Co., Ltd., the development of the fuel transfer machines which take out and charge the fuel for the reactors and carry out the transport between fuel storage facilities is important subject. In this paper, the outline of the development of a fuel transfer machine is described, centering around the handling of fuel in sodium and the removal of the heat of spent fuel. ''Monju'' with 300 MW electric output is being developed for the purposes of putting fast breeder reactors in practical use, increasing plant power output, improving the rate of operation, and securing the safety. Plutonium is used as the fuel, and liquid metallic sodium is used as the coolant in fast breeder reactors, accordingly the fuel must be handled safely in gas-tight, shielded vessels while it is cooled, and securely in sodium and sodium cover gas by remote operation. Gripper, gripper-driving mechanism, coffin, movable block, door valves, cooling system and carriage compose the fuel transfer machine, and these are described. The main results of development and the tests for the development, such as the trial manufacture of gripper and decay heat removal test, are reported. (Kako, I.)

  4. Average cross sections calculated in various neutron fields

    Shibata, Keiichi

    2002-01-01

    Average cross sections have been calculated for the reactions contained in the dosimetry files, JENDL/D-99, IRDF-90V2, and RRDF-98 in order to select the best data for the new library IRDF-2002. The neutron spectra used in the calculations are as follows: 1) 252 Cf spontaneous fission spectrum (NBS evaluation), 2) 235 U thermal fission spectrum (NBS evaluation), 3) Intermediate-energy Standard Neutron Field (ISNF), 4) Coupled Fast Reactivity Measurement Facility (CFRMF), 5) Coupled thermal/fast uranium and boron carbide spherical assembly (ΣΣ), 6) Fast neutron source reactor (YAYOI), 7) Experimental fast reactor (JOYO), 8) Japan Material Testing Reactor (JMTR), 9) d-Li neutron spectrum with a 2-MeV deuteron beam. The items 3)-7) represent fast neutron spectra, while JMTR is a light water reactor. The Q-value for the d-Li reaction mentioned above is 15.02 MeV. Therefore, neutrons with energies up to 17 MeV can be produced in the d-Li reaction. The calculated average cross sections were compared with the measurements. Figures 1-9 show the ratios of the calculations to the experimental data which are given. It is found from these figures that the 58 Fe(n, γ) cross section in JENDL/D-99 reproduces the measurements in the thermal and fast reactor spectra better than that in IRDF-90V2. (author)

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

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

    1989-01-01

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

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

    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.)

  7. High temperature structure design for FBRs and analysis technology

    Iwata, Koji

    1986-01-01

    In the case of FBRs, the operation temperature exceeds 500 deg C, therefore, the design taking the inelastic characteristics of structural materials, such as plasticity and creep, into account is required, and the high grade and detailed evaluation of design is demanded. This new high temperature structure design technology has been advanced in respective countries taking up experimental, prototype and demonstration reactors as the targets. The development of FBRs in Japan was begun with the experimental reactor 'Joyo' which has been operated since 1977, and now, the prototype FBR 'Monju' of 280 MWe is under construction, which is expected to attain the criticality in 1992. In order to realize FBRs which can compete with LWRs through the construction of a demonstration FBR, the construction of large scale plants and the heightening of the economy and reliability are necessary. The features and the role of FBR structural design, the method of high temperature structure design and the trend of its standardization, the trend of the structural analysis technology for FBRs such as inelastic analysis, buckling analysis and fluid and structure coupled vibration analysis, the present status of structural analysis programs, and the subjects for the future of high temperature structure design are explained. (Kako, I.)

  8. A review of fast reactor programme in Japan

    Matsuno, Y.; Bando, S.

    1981-03-01

    The fast breeder reactor development project in Japan made progress in the past year, and will be continued in the next fiscal 1981. The scale of efforts both in budget and personnel will be similar to those in fiscal 1980. The budget for R and D works and for the construction of the fast breeder prototype reactor ''Monju'' will be approximately 20 billion yen and 27 billion yen, respectively, excluding the wage of the personnel concerned. The number of the technical personnel currently engaging in fast breeder reactor development in the Power Reactor and Nuclear Fuel Development Corp. is about 530. As for the experimental fast reactor ''Joyo'', three operational cycles at 75 MWt have been completed in August, 1980, and the fourth cycle has started in March, 1981. As for the prototype reactor ''Monju'', progress was made toward the construction, and the environmental impact statement on the reactor was approved by the authorities concerned. The studies on the preliminary design of large LMFBRs have been made by the PNC and also by power companies. The design study carried out by the PNC is concerned with a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of the commissioning of ''Monju''. The highlights and topics in the development activities for fast breeder reactors in the past twelve months are summarized in this report. (Kako, I.)

  9. Simplified simulation of an experimental fast reactor plant

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  10. Integral test of JENDL-3.3 on fast reactors

    Chiba, Gou; Hazama, Taira

    2003-05-01

    An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

  11. Construction of 'Monju' to begin this year

    1980-01-01

    In March, 1980, the Atomic Energy Commission, Japan, modified its basic policy on fast breeder reactors, which is the first modification in twelve years. It postponed the target time of commercializing FBRs by ten years, in accordance with the present level of the development, from the original target of 1985. But the construction of the prototype FBR ''Monju'' will be started this year in Fukui Prefecture as the successor to the experimental FBR ''Joyo'' of 75 MWt. The attainment of criticality is expected around 1987. The percentage of the construction cost to be borne by the electric utilities and manufacturers was decided at 20% by the AEC as it increased to some 400 billion yen do to inflation. In the budget for FY 1980, the government has set aside 15.6 billion yen for the construction of ''Monju''. The environmental assessment concerning the planned construction of ''Monju'' in Tsuruga City, submitted by the PNC, was approved by the Fukui Prefectural Council for the Preservation of Natural Environment. The construction will be approved by the end of 1980 by the governor of Fukui. ''Monju'' is a sodium-cooled fast neutron reactor of loop type with 300 MWe output. Its design has been continued from 1968 to 1976. The outline of ''Monju'', its core and fuel, the equipment for the cooling system, the steam generators, the treatment of radioactive wastes and buildings are described. (Kako, I.)

  12. Establishment of technological basis for fabrication of U-Pu-Zr ternary alloy fuel pins for irradiation tests in Japan

    Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Arai, Yasuo; Nakamura, Kinya; Ogata, Takanari

    2011-01-01

    A high-purity Ar gas atmosphere glove box accommodating injection casting and sodium-bonding apparatuses was newly installed in the Plutonium Fuel Research Facility of Oarai Research and Development Center, Japan Atomic Energy Agency, in which several nitride and carbide fuel pins were fabricated for irradiation tests. The experiences led to the establishment of the technological basis of the fabrication of U-Pu-Zr alloy fuel pins for the first time in Japan. After the injection casting of the U-Pu-Zr alloy, the metallic fuel pins were fabricated by welding upper and lower end plugs with cladding tubes of ferritic-martensitic steel. Subsequent to the sodium bonding for filling the annular gap region between the U-Pu-Zr alloy and the cladding tube with the melted sodium, the fuel pins for irradiation tests are inspected. This paper shows the apparatuses and the technological basis for the fabrication of U-Pu-Zr alloy fuel pins for the irradiation test planned at the experimental fast test reactor Joyo. (author)

  13. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  14. Acoustic Leak Detection Requirements for a SFR Steam Generator Protection

    Kim, Tae-Joon; Jeong, Ji-Young; Oeh, Jae-Hyuk; Kim, Jong-Man; Kim, Byung-Ho; Yughay, Valery S.

    2008-01-01

    A large volume of fast reactor research has been executed in Russia, Japan, France, India and the United Kingdom. At present, an unique fast reactor named BN- 600 is operating in Russia. Also, the operation of research reactors such as Phenix (France), JOYO (Japan), BOR-60 (Russia) and FBTR (India) proceeds. The last project to be completed was the reactor Monju (Japan) which is now stopped. In addition activities for the development of fast reactors are being conducted in China, India, and South Korea. Fast reactors are a choice for the subsequent nuclear power generation in Korea, and their increased safety is one of the basic requirements. The basis for a tightening of the requirements on safety is the emergencies in NPPs in Russia, USA, France, Japan and other countries. These emergencies testify that the existing monitoring systems do not fully provide a well-timed detection of the distresses arising in a NPP, because of a poor sensitivity and response, thus the necessity for a better diagnostic system is obvious. In accordance with the USA GNEP initiative in Obninsk, Russia, 2007 the main efforts should be directed toward a sodium-water steam generator safety increase due to improvement of the hydrogen monitoring system and the acoustic leak detection system

  15. Evaluation of remaining behavior of halogen on the fabrication of MOX pellet containing Am

    Ozaki, Yoko; Osaka, Masahiko; Obayashi, Hiroshi; Tanaka, Kenya

    2004-11-01

    It is important to limit the content of halogen elements, namely fluorine and chlorine that are sources of making cladding material corrode, in nuclear fuel from the viewpoint of quality assurance. The halogen content should be more carefully limited in the MOX fuel containing Americium (Am-MOX), which is fabricated in the Alpha-Gamma Facility (AGF) for irradiation testing to be conducted in the experimental fast reactor JOYO, because fluorine may remain in the sintered pellets owing to a formation of AmF 3 known to have a low vapor pressure and may exceeds the limit of 25 ppm. In this study, a series of experimental determination of halogen element in Am-MOX were performed by a combination method of pyrolysis and ion-chromatography for the purpose of an evaluation of behavior of remaining halogen through the sintering process. Oxygen potential, temperature and time were changed as experimental parameters and their effects on the remaining behavior of halogen were examined. It was confirmed that good pellets, which contained small amount of halogen, could be obtained by the sintering for 3 hour at 1700degC in the oxygen potential range from -520 to -390 kJ/mol. In order to analysis of fluorine chemical form in green pellet, thermal analysis was performed. AmF 3 and PuF 3 have been confirmed to remain in the green pellet. (author)

  16. A review of fast reactor program in Japan

    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

  17. A review of fast reactor program in Japan

    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

  18. Development of an electrical connector for liquid sodium environment. Final Report

    Kataoka, Hajime; Noguchi, Koichi; Takatsudo, Hiroshi; Miyakawa, Shun-ichi

    1998-07-01

    The INstrumented irradiation Test Assembly (INTA) has been used to conduct precision on-line instrumented irradiation tests in the experimental fast reactor JOYO. In INTA, direct instrumentation wiring between the irradiation test section in the core and the upper structure section in the rotating plug makes INTA structurally complex and expensive. Instead of direct wiring, if an electrical connector capable of withstanding a heated liquid sodium environment could be used between the irradiation test section and the upper structure section, the upper mechanism of INTA could be reused and testing costs would be drastically reduced. Moreover, the reactor load factor would be improved because of reduced handling time for INTA. In an attempt to gain this advantage, research and development of an electric connector in a sodium environment was carried out from 1988 to 1996 at PNC. As no previous R and D had been conducted in this area, this development activity was conducted in a boot strap manner. The first test was carried out for a small model fabrication, the second was for a water partial model, and the third was for a sodium partial model. Based on those tests, a prototype design specification of the connector was determined. In the sodium partial model test, the resilience of the electrical connector insulation to the sodium environment was investigated. However, severe cracking in the ceramic insulator caused by the high temperature sodium environment was discovered at the junction of ceramic insulator and metallic electrode. Although additional sodium partial tests were performed for various material combinations of ceramic insulators, metallic electrodes, brazing materials and metallization materials, the results of the tests were unsatisfactory. Therefore, it was decided that the development of the connector in sodium should cease at PNC in 1997. (J.P.N.)

  19. Proceedings of the third JAERI-KAERI joint seminar on post irradiation examination technology

    1999-09-01

    Between the Department of JMTR of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Cycle Research Group of the Korea Atomic Energy Research Institute (KAERI), it has been periodically carried out the collaboration on technical information exchange by specialists and scientists, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI. And JAERI-KAERI joint seminar has been held every three years. The 1st and 2nd JAERI-KAERI Joint Seminars were held in November 1992 at JAERI and in September 1995 at KAERI, respectively. The 3rd JAERI-KAERI Joint Seminar was held on 25 and 26 March, 1999 at the Oarai Research Establishment of JAERI. In this seminar, total participants of 84 were joined from JAERI, KAERI, Hanyang University, Japan Nuclear Cycle Development Institute, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. Contributed presentations were in three sessions; Current status and future perspectives on PIE, PIE techniques and Evaluation of PIE data. Re-assembling technique for JOYO fuel, Nd-YAG laser welding technique, grain boundary analysis using FEG-TEM, lift time estimation of PWR Rod Cluster Control Assembly (RCCA) rodlet and failure analysis of Korea Nuclear Power Plant (KNP) fuel have been widely noticed as topic items on PIE. And some comments from PIE user, were pointed out that the nano-PIE technique, the flexibility to ad-hoc demands on testing space or utilization, and the international collaboration were very important for the next generation's PIE. The 34 of the present papers are indexed individually. (J.P.N.)

  20. A review of fast reactor programme in Japan

    Masuno, Y [Experimental Fast Reactor Division, O-arai Engineering Center, PNC (Japan); Bando, S [Project Planning and Management Division, PNC, Minato-ku, Tokyo (Japan)

    1981-05-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report.

  1. Sophistication of burnup analysis system for fast reactor

    Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

    2010-02-01

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  2. Conceptual design of the alcohol waste treatment equipment

    Fujisawa, Morio; Nitta, Kazuhiko; Morita, Yasuhiro; Nakada, Eiju

    2001-01-01

    This report describes the result of Conceptual Design of the Alcohol Waste Treatment Equipment. The experimental fast Reactor, JOYO, saves the radioactive alcohol waste at storage tank. As this alcohol waste is not able to treat with existing equipment, it is stored about 5 m 3 . And the amount of this is increasing every year. So it is necessary to treat the alcohol waste by chemical resolution for example. On account of this, the investigative test about filtration and dialyzer, and conceptual design about catalyst oxidation process, which is composed from head end process to resolution, are done. The results of investigation show as follows. 1. Investigative Test about filtration and dialyzer. (1) The electric conduction is suitable for the judgement of alkyl sodium hydrolysis Alkyl sodium hydrolysis is completed below 39% alcohol concentration. (2) The microfiltration is likely to separate the solid in alcohol waste. (3) From laboratory test, the electrodialyzer is effective for sodium separation in alcohol waste. And sodium remove rate, 96-99%, is confirmed. 2. Conceptual Design. The candidate process is as follows. (1) The head end process is electrodialyzer, and chemical resolution process is catalyst oxidation. (2) The head end process is not installed, and chemical resolution process is catalyst oxidation. (3) The head end process is electrodialyzer, and alcohol extracted by pervaporation. In this Conceptual Design, as far these process, the components, treatment ability, properties of waste, chemical mass balance, safety for fire and explosion, and the plot plan are investigated. As a result, remodeling the existing facility into catalyst oxidation process is effective to treat the alcohol waste, and treatment ability is about 1.25 l/h. (author)

  3. Development Status on Innovative Sodium-Cooled Fast Reactor (JSFR)

    Yanagisawa, Tsutomu; Sato, Kazujiro

    2006-01-01

    The first step in Japan's nuclear fuel cycle policy is to introduce MOX recycle in light water reactors (LWRs) and the final step is to establish multiple TRU recycle in fast reactors (FRs), with the goal of realizing a stable supply, effective use of nuclear fuel resources, and the environmentally friendly production of energy. Therefore, a feasibility study on commercialized FR cycle systems has been launched since July 1999 by a Japanese joint project team of Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC: the representative of the electric utilities) in cooperation with Central Research Institute of Electric Power Industry (CRIEPI) and vendors. In the period from July 1999 to March 2001, the feasibility study phase-I was conducted to screen out representative FR cycle concepts. In the feasibility study phase-II (April 2001 - March 2006), investigations in to the representative FR concepts were carried out to clarify the most promising concept for commercial deployment. This paper describes an innovative sodium-cooled FR, which is named as the JAEA Sodium-cooled FR (JSFR), as the most promising FR concept that meets the Generation-IV performance target. The JSFR employs several advanced technologies, such as an oxide dispersion strengthened (ODS) cladding for higher burn-up, a short-piping configuration with less elbows by adopting high chromium steel, a large scale integrated intermediate heat exchanger with a primary circulation pump, etc. Based on the design, construction and operation experiences of JOYO and MONJU, there are extensive technology bases for sodium-cooled FRs. Nevertheless, several innovative technologies implemented into the JSFR have to be developed in order to realize higher economic competitiveness by reducing construction costs and improving plant availability

  4. Fast critical experiments in FCA and their analysis

    Hirota, Jitsuya

    1984-02-01

    JAERI Fast Critical Facility FCA went critical for the first time in April, 1967. Since then, critical experiments and their analysis were carried out on thirty-five assemblies until march, 1982. This report summarizes many achievements obtained in these fifteen years and points out disagreements observed between the calculation and experiment for further studies. A series of mock-up experiments for Experimental Fast Reactor JOYO, a theoretical and numerical study of adjustment of group constants by using integral data and a development of proton-recoil counter system for fast neutron spectrum measurement won high praise. Studies of Doppler effect of structural materials, effect of fission product accumulation on sodium-void worth, axially heterogeneous core and actinide cross sections attracted world-side attention. Significant contributions were also made to Prototype Fast Breeder Reactor MONJU through the partial mock-up experiments. Disagreements between the calculation and experiment were observed in the following items; reaction rate distribution and reactivity worth of B 4 C absorber in radial blanket, central reactivity worth in core with reflector, plate/pin fuel heterogeneity effect on criticality, sodium-void effect in central core region, Doppler effect of structural materials, core neutron spectrum near large resonances of iron and oxygen, effect of fission product accumulation on sodium-void worth, physics property of heterogeneous core, reactivity change resulted from fuel slumping and so on. Further efforts should be made to solve these disagreements through recalculating the experimental results with newly developed data and methods and carrying out the experiments intended to identify the cause of disagreement. (author)

  5. Gas Test Loop Facilities Alternatives Assessment Report Rev 1

    William J. Skerjanc; William F. Skerjanc

    2005-01-01

    An important task in the Gas Test Loop (GTL) conceptual design was to determine the best facility to serve as host for this apparatus, which will allow fast-flux neutron testing in an existing nuclear facility. A survey was undertaken of domestic and foreign nuclear reactors and accelerator facilities to arrive at that determination. Two major research reactors in the U.S. were considered in detail, the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), each with sufficient power to attain the required neutron fluxes. HFIR routinely operates near its design power limit of 100 MW. ATR has traditionally operated at less than half its design power limit of 250 MW. Both of these reactors should be available for at least the next 30 years. The other major U.S. research reactor, the Missouri University Research Reactor, does not have sufficient power to reach the required neutron flux nor do the smaller research reactors. Of the foreign reactors investigated, BOR-60 is perhaps the most attractive. Monju and BN 600 are power reactors for their respective electrical grids. Although the Joyo reactor is vigorously campaigning for customers, local laws regarding transport of radioactive material mean it would be very difficult to retrieve test articles from either Japanese reactor for post irradiation examination. PHENIX is scheduled to close in 2008 and is fully booked until then. FBTR is limited to domestic (Indian) users only. Data quality is often suspect in Russia. The only accelerator seriously considered was the Fuel and Material Test Station (FMTS) currently proposed for operation at Los Alamos National Laboratory. The neutron spectrum in FMTS is similar to that found in a fast reactor, but it has a pronounced high-energy tail that is atypical of fast fission reactor spectra. First irradiation in the FMTS is being contemplated for 2008. Detailed review of these facilities resulted in the recommendation that the ATR would be the best host for the GTL

  6. A review of fast reactor program in Japan - April 1983

    Matsuno, Y.

    1983-01-01

    The fast breeder reactor development project in Japan has been in progress during the past twelve months and will be continued in the next fiscal year, from April 1983 through March 1984, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1982. Concerning the experimental fast reactor, JOYO, all the scheduled testings and operations were completed by the end of 1981 and from the beginning of 1982 the change-out work from Mark-I core to Mark-II core has been continued for 11 months. The initial criticality on the Mark-II core was achieved on 22 Nov. 1982 and after 3 months low power physics tests the reactor power was raised up to 100% (100 MWt) in the middle of March 1983. With respect to the prototype reactor MONJU, progress toward construction has been made and the licensing of the second step will be completed in the first half of 1983. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  7. Actinide consumption: Nuclear resource conservation without breeding

    Hannum, W.H.; Battles, J.E.; Johnson, T.R.; McPheeters, C.C.

    1991-01-01

    A new approach to the nuclear power issue based on a metallic fast reactor fuel and pyrometallurgical processing of spent fuel is showing great potential and is approaching a critical demonstration phase. If successful, this approach will complement and validate the LWR reactor systems and the attendant infrastructure (including repository development) and will alleviate the dominant concerns over the acceptability of nuclear power. The Integral Fast Reactor (IFR) concept is a metal-fueled, sodium-cooled pool-type fast reactor supported by a pyrometallurgical reprocessing system. The concept of a sodium cooled fast reactor is broadly demonstrated by the EBR-II and FFTF in the US; DFR and PFR in the UK; Phenix and SuperPhenix in France; BOR-60, BN-350, BN-600 in the USSR; and JOYO in Japan. The metallic fuel is an evolution from early EBR-II fuels. This fuel, a ternary U-Pu-Zr alloy, has been demonstrated to be highly reliable and fault tolerant even at very high burnup (160-180,000 MWd/MT). The fuel, coupled with the pool type reactor configuration, has been shown to have outstanding safety characteristics: even with all active safety systems disabled, such a reactor can survive a loss of coolant flow, a loss of heat sink, or other major accidents. Design studies based on a small modular approach show not only its impressive safety characteristics, but are projected to be economically competitive. The program to explore the feasibility of actinide recovery from spent LWR fuel is in its initial phase, but it is expected that technical feasibility could be demonstrated by about 1995; DOE has not yet committed funds to achieve this objective. 27 refs.

  8. Studies on Enhancing Nuclear Transparency in the Asia-Pacific Region

    Kawakubo, Y.; Tomikawa, H.

    2015-01-01

    Nuclear transparency is defined as ''a cooperative process of providing information to all interested parties so that they can independently assess the safety, security, and legitimate management of nuclear materials'' by Sandia National Laboratories (SNL). Since the Asia- Pacific region has a broad spectrum of nuclear development underway and planned in the future, nuclear transparency is recognized as essential to provide additional assurance and enhance confidence building in this area. It is expected that elevated nuclear transparency should also supplement International Atomic Energy Agency (IAEA) safeguards. With this recognition, JAEA has committed various studies and activities for enhancing regional nuclear transparency mainly with U.S. Department of Energy (DOE) and its national laboratories. The efforts include concept study, development of secure data transmission technologies at the Experimental Fast Reactor ''Joyo'' for the use of regional nuclear transparency, and support for Council for Security and Cooperation in Asia Pacific (CSCAP) to develop internet-based transparency tools. JAEA also organized several workshops to discuss with stakeholder organizations to build acceptance for transparency tools and activities. Based on the past studies, JAEA, jointly with SNL, Korea Institute of Nuclear Nonproliferation and Control (KINAC) and Korea Atomic Energy Research Institute (KAERI), initiated a new phase of study in 2011 to design and establish an Information Sharing Framework (ISF) which was defined as ''a communication platform on which nuclear nonproliferation experts can provide and/or receive relevant information in a practical and sustainable manner''. During the period of two-year study, partner organizations identified essential elements to establish ISF and developed the requirements. Currently, JAEA and KINAC are planning to implement demonstration of ISF under Asia Pacific

  9. Proceedings of the third JAERI-KAERI joint seminar on post irradiation examination technology

    NONE

    1999-09-01

    Between the Department of JMTR of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Cycle Research Group of the Korea Atomic Energy Research Institute (KAERI), it has been periodically carried out the collaboration on technical information exchange by specialists and scientists, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI. And JAERI-KAERI joint seminar has been held every three years. The 1st and 2nd JAERI-KAERI Joint Seminars were held in November 1992 at JAERI and in September 1995 at KAERI, respectively. The 3rd JAERI-KAERI Joint Seminar was held on 25 and 26 March, 1999 at the Oarai Research Establishment of JAERI. In this seminar, total participants of 84 were joined from JAERI, KAERI, Hanyang University, Japan Nuclear Cycle Development Institute, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. Contributed presentations were in three sessions; Current status and future perspectives on PIE, PIE techniques and Evaluation of PIE data. Re-assembling technique for JOYO fuel, Nd-YAG laser welding technique, grain boundary analysis using FEG-TEM, lift time estimation of PWR Rod Cluster Control Assembly (RCCA) rodlet and failure analysis of Korea Nuclear Power Plant (KNP) fuel have been widely noticed as topic items on PIE. And some comments from PIE user, were pointed out that the nano-PIE technique, the flexibility to ad-hoc demands on testing space or utilization, and the international collaboration were very important for the next generation's PIE. The 34 of the present papers are indexed individually. (J.P.N.)

  10. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan); Ito, Masahiro; Maeda, Koji [Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki 311-1393 (Japan)

    2011-09-30

    Highlights: > We evaluated diametral strain of fast reactor MOX fuel pins irradiated to 130 GWd/t. > The strain was due to cladding void swelling and irradiation creep. > The irradiation creep was caused by internal gas pressure and PCMI. > The PCMI was associated with pellet swelling by rim structure or by cesium uranate. > The latter effect tended to increase the cumulative damage fraction of the cladding. - Abstract: The C3M irradiation test, which was conducted in the experimental fast reactor, 'Joyo', demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, 'Monju'. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and {sup 137}Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  11. Outline of facility for studying high level radioactive materials (CPF) and study programmes

    Sakamoto, Motoi

    1983-01-01

    The Chemical Processing Facility for studying high level radioactive materials in Tokai Works of Power Reactor and Nuclear Fuel Development Corp. is a facility for fundamental studies centering around hot cells, necessary for the development of fuel recycle techniques for fast breeder reactors, an important point of nuclear fuel cycle, and of the techniques for processing and disposing high level radioactive liquid wastes. The operation of the facility was started in 1982, for both the system A (the test of fuel recycle for fast breeder reactors) and the system B (the test of vitrification of high level liquid wastes). In this report, the outline of the facility, the contents of testings and the reflection of the results are described. For the fuel recycle test, the hot test of the spent fuel pins of JOYO MK-1 core was started, and now the uranium and plutonium extraction test is underway. The scheduled tests are fuel solubility, the confirmation of residual properties in fuel melting, the confirmation of extracting conditions, the electrolytic reduction of plutonium, off-gas behaviour and the test of material reliability. For the test of vitrification of high level liquid wastes, the fundamental test on the solidifying techniques for the actual high level wastes eluted from the Tokai reprocessing plant has been started, and the following tests are programmed: Assessment of the properties of actual liquid wastes, denitration and concentration test, vitrification test, off-gas treatment test, the test of evaluating solidified wastes, and the test of storing solidified wastes. These test results are programmed to be reflected to the safety deliberation and the demonstration operation of a vitrification pilot plant. (Wakatsuki, Y.)

  12. Development of decommissioning management system. 9. Remodeling to PC system and system verification by evaluation of real work

    Kondo, Hitoshi; Fukuda, Seiji; Okubo, Toshiyuki

    2004-03-01

    When the plan of decommissioning such as nuclear fuel cycle facilities and small-scale research reactors is examined, it is necessary to select the technology and the process of the work procedure, and to optimize the index (such as the radiation dose, the cost, amount of the waste, the number of workers, and the term of works, etc.) concerning dismantling the facility. In our waste management section, Development of the decommissioning management system, which is called 'DECMAN', for the support of making the decommissioning plan is advanced. DECMAN automatically calculates the index by using the facility data and dismantling method. This paper describes the remodeling of program to the personal computer and the system verification by evaluation of real work (Dismantling of the liquor dissolver in the old JOYO Waste Treatment Facility (the old JWTF), the glove boxes in Deuterium Critical Assembly (DCA), and the incinerator in Waste Dismantling Facility (WDF)). The outline of remodeling and verification is as follows. (1) Additional function: 1) Equipment arrangement mapping, 2) Evaluation of the radiation dose by using the air dose rate, 3) I/O of data that uses EXCEL (software). (2) Comparison of work amount between calculation value and results value: The calculation value is 222.67man·hour against the result value 249.40 man·hour in the old JWTF evaluation. (3) Forecast of accompanying work is predictable to multiply a certain coefficient by the calculation value. (4) A new idea that expected the amount of the work was constructed by using the calculation value of DECMAN. (author)

  13. A review of fast reactor progress in Japan, March 1979

    Tomabechi, K

    1979-07-01

    The fast reactor development project in Japan will be continued in the next fiscal year, from April 1979 through March 1980, at a similar scale of effort both in budget and personnel, to those of the fiscal year of 1978. The total budget for LMFBR development for the next fiscal year is approximately 24 billion Yen, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast reactor development in the PNC is approximately 500, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, approval for power increase from presently approved 50 MWt to 75 MWt with the present core and also to 100 MWt with a modified core in the future was granted by the regulatory authority in September 1978. Two operational cycles at 50 MWt have been completed very recently and preparation for power increase to 75 MWt is being made. With respect to the prototype fast breeder reactor MONJU, progress toward construction is being made and an environmental impact statement of MONJU filed last autumn is being reviewed by the concerned authorities. By the new atomic energy law recently made effective in Japan, the tasks of the former Japan Atomic Energy Commission were split into two and the Atomic Energy Safety Commission was newly established on 4th October 1978 in order to deal with nuclear safety problems in the country. All other problems are treated by the Atomic Energy Commission, as before. Highlights and topics of the fast reactor development activities in the past twelve months are summarized in this paper.

  14. Basic plan of partitioning and transmutation technology development

    Ikegami, Tetsuo; Ozawa, Masaki

    2003-04-01

    Basic plan of partitioning and transmutation technology development has been made in more detail and concrete manner in terms of development goal, nuclides to be portioned and to be transmuted, and development schedule, based on the pre-evaluation results of the Research Evaluation Committee on Research and development of partitioning and transmutation technology for long life nuclides' held in August 2000. A step by step approach, consists of three steps, to reach the goal of partitioning and transmutation technology has been adopted under the recognition that the partitioning and transmutation technology development should be progressed steadily as a long term them. The first step is supposed to be able to attain within about 5 years by the present technology and on the extension of it. Such researches as collective separation of TRU, MA/Ln effective separation, and irradiation experiment of iodine and technetium. The second step is such a goal that is expected to be able to realize the engineering feasibility, within about 15 years, through the progress of science technology in future, although the engineering feasibility is not sufficiently foreseen at present. It will need revolutionary technology or breakthrough. Nuclides to be partitioned and to be transmuted have been selected in view points of 'radioactivity and radio-toxicity', 'geological repository', and 'effective utilization', corresponding to the each step of the development goal. Collaboration with other research organizations and with universities in the world should be pursued. Especially, such collaborations with France, with which information exchange on JOYO/PHENIX irradiation experiments is progressing, and with USA, which has recently developed positive activities in this field, are strongly expected. (author)

  15. A review of fast reactor programme in Japan

    Masuno, Y.; Bando, S.

    1981-01-01

    The fast breeder reactor development project in Japan has been in progress in the past twelve months and will be continued in the next fiscal year, from April 1981 through March 1982, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1980. The 1981 year budget for P and D work and for construction of a prototype fast breeder reactor, Monju, will be approximately 20 and 27 billion Yen respectively, excluding wages of the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaging in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor, Joyo, power increase from 50 MWt to 75 MWt was made in July 1979 and three operational cycles at 75 MWt have been completed in August 1980 and the forth cycle has started in the middle of March 1981. With respect to the prototype reactor Monju, progress toward construction has been made and an environmental impact statement of the reactor was approved by the concerned authorities. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MW e plant of loop type by extrapolating the technology to be developed by the time of commissioning of Monju. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor Monju. Highlights and topics of the fast breeder reactor development activities in the past twelve months are summarized in this report

  16. A review of fast reactor program in Japan

    Matsuno, Y.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress for the past twelve months and will be continued this fiscal year, from April 1982 through March 1983, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1981. The 1982 year budget for R and D work and for construction of a prototype fast breeder reactor MONJU is approximately 20 and 27 billion yen respectively, excluding wages for the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaged in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and six operational cycles at 75 MWt were completed in December 1981. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor was approved by the authorities concerned, and the licensing of the first step was completed at the end of 1981. Preliminary design studies of a large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  17. The Effects of Internet Shoppers' Trust on their purchasing intention in China

    Rong Li

    2007-12-01

    Full Text Available Owing to the rapid development of the Internet and information technology in China, the growth of consumers’ purchasing activities in Internet shopping malls has been truly phenomenal in recent years. Taobao.com, Ebay.com.cn, and Paipai.com have 67,360,000 customer to customer (C2C users and 99% of the market share in China’s C2C market (www.163.com. Dangdang.com and Joyo.com have occupied 87% of the business to customer (B2C market with 58,360,000 users (www.sohu.com. Because of these significant numbers of users, it is important to understand what affects Chinese consumers’ decisions to purchase in Internet shopping malls. Based on past studies, trust is considered a key factor affecting a Chinese consumer’s purchasing intention. The purpose of this study is to investigate the effects of Chinese shoppers’ trust on their purchasing intention in Internet shopping malls. In order to accomplish the purpose of this study, we developed a research model. This model suggests that there exists a significant relationship between trust and purchasing intention. According to this model, on purchasing intention, trust also mediates effects of other independent variables such as e-commerce knowledge, perceived reputation, perceived risk, and perceived ease of use. The results of this study show that the relationships between these variables are all significant except that between trust and perceived reputation. This research confirms the significant effects of Chinese shoppers’ trust on purchasing intention. Implications of these findings are discussed for researchers and practitioners.

  18. Investigation of the fuel temperature evaluation method at BOL

    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)

  19. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  20. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

    2010-01-31

    An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

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

    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)

  2. Actinide consumption: Nuclear resource conservation without breeding

    Hannum, W.H.; Battles, J.E.; Johnson, T.R.; McPheeters, C.C.

    1991-01-01

    A new approach to the nuclear power issue based on a metallic fast reactor fuel and pyrometallurgical processing of spent fuel is showing great potential and is approaching a critical demonstration phase. If successful, this approach will complement and validate the LWR reactor systems and the attendant infrastructure (including repository development) and will alleviate the dominant concerns over the acceptability of nuclear power. The Integral Fast Reactor (IFR) concept is a metal-fueled, sodium-cooled pool-type fast reactor supported by a pyrometallurgical reprocessing system. The concept of a sodium cooled fast reactor is broadly demonstrated by the EBR-II and FFTF in the US; DFR and PFR in the UK; Phenix and SuperPhenix in France; BOR-60, BN-350, BN-600 in the USSR; and JOYO in Japan. The metallic fuel is an evolution from early EBR-II fuels. This fuel, a ternary U-Pu-Zr alloy, has been demonstrated to be highly reliable and fault tolerant even at very high burnup (160-180,000 MWd/MT). The fuel, coupled with the pool type reactor configuration, has been shown to have outstanding safety characteristics: even with all active safety systems disabled, such a reactor can survive a loss of coolant flow, a loss of heat sink, or other major accidents. Design studies based on a small modular approach show not only its impressive safety characteristics, but are projected to be economically competitive. The program to explore the feasibility of actinide recovery from spent LWR fuel is in its initial phase, but it is expected that technical feasibility could be demonstrated by about 1995; DOE has not yet committed funds to achieve this objective. 27 refs

  3. Recovery of minor actinides from spent fuel using TPEN-immobilized gels

    Koyama, S.; Suto, M.; Ohbayashi, H.; Oaki, H.; Takeshita, K.

    2013-01-01

    A series of separation experiments was performed in order to study the recovery process for minor actinides (MAs), such as americium (Am) and curium (Cm), from the actual spent fuel by using an extraction chromatographic technique. N,N,N',N'-tetrakis-(4-propenyloxy-2-pyridylmethyl) ethylenediamine (TPPEN) is an N,N,N',N'-tetrakis (2-pyridylmethyl) ethylenediamine (TPEN) analogue consisting of an incorporated pyridine ring that acts as not only a ligand but also as a site for polymerization and crosslinking of the gel. The TPPEN and N-isopropylacrylamide (NIPA) were dissolved into dimethylformamide (DMF, Wako Co., Ltd.) and a silica beads polymer, and then TTPEN was immobilized chemically in a polymer gel (so called TPEN-gel). Mixed oxide (MOX) fuel, which was highly irradiated up to 119 GWD/MTM in the experimental fast reactor Joyo, was used as a reference spent fuel. First, uranium (U) and plutonium (Pu) were separated from the irradiated fuel using an ion-exchange method, and then, the platinum group elements were removed by CMPO to leave a mixed solution of MAs and lanthanides. The 3 mol% TPPEN-gel was packed with as an extraction column (CV: 1 ml) and then rinsed by 0.1 M NaNO 3 (pH 4.0) for pH adjustment. After washing the column by 0.01 M NaNO 3 (pH 4.0), Eu was detected and the recovery rate reached 93%. The MAs were then recovered by changing the eluent to 0.01 M NaNO 3 (pH 2.0), and the recovery rate of Am was 48 %. The 10 mol% TPPEN-gel was used to improve adsorption coefficient of Am and a condition of eluent temperature was changed in order to confirm the temperature swing effect on TPEN-gel for MA. More than 90% Eu was detected in the eluent after washing with 0.01 M NaNO 3 (pH 3.5) at 5 Celsius degrees. Americium was backwardly detected and eluted continuously during the same condition. After removal of Eu, the eluent temperature was changed to 32 Celsius degrees, then Am was detected (pH 3.0). Finally remained Am could be stripped from TPPEN-gel by

  4. Systemization of burnup sensitivity analysis code

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  5. Application of statistical method for FBR plant transient computation

    Kikuchi, Norihiro; Mochizuki, Hiroyasu

    2014-01-01

    Highlights: • A statistical method with a large trial number up to 10,000 is applied to the plant system analysis. • A turbine trip test conducted at the “Monju” reactor is selected as a plant transient. • A reduction method of trial numbers is discussed. • The result with reduced trial number can express the base regions of the computed distribution. -- Abstract: It is obvious that design tolerances, errors included in operation, and statistical errors in empirical correlations effect on the transient behavior. The purpose of the present study is to apply above mentioned statistical errors to a plant system computation in order to evaluate the statistical distribution contained in the transient evolution. A selected computation case is the turbine trip test conducted at 40% electric power of the prototype fast reactor “Monju”. All of the heat transport systems of “Monju” are modeled with the NETFLOW++ system code which has been validated using the plant transient tests of the experimental fast reactor Joyo, and “Monju”. The effects of parameters on upper plenum temperature are confirmed by sensitivity analyses, and dominant parameters are chosen. The statistical errors are applied to each computation deck by using a pseudorandom number and the Monte-Carlo method. The dSFMT (Double precision SIMD-oriented Fast Mersenne Twister) that is developed version of Mersenne Twister (MT), is adopted as the pseudorandom number generator. In the present study, uniform random numbers are generated by dSFMT, and these random numbers are transformed to the normal distribution by the Box–Muller method. Ten thousands of different computations are performed at once. In every computation case, the steady calculation is performed for 12,000 s, and transient calculation is performed for 4000 s. In the purpose of the present statistical computation, it is important that the base regions of distribution functions should be calculated precisely. A large number of

  6. Research and development of pyro-reprocessing and its world status

    Inoue, Tadashi

    2005-01-01

    The next generation fuel cycle requires a strong resistance of nuclear proliferation and lightening the environmental burden as well as safety and economic advantage. The pyro-reprocessing technology satisfies such kinds of requirements. Central Research Institute of Electric Power Industry, CRIEPI, has been involving the development of metal fuel cycle integrated pyro-reprocessing with metal-electrorefining and metal fuel fast reactor since 1986. The study on pyro-processing technology of spent MOX fuel from LWR has been also started. Based on the fast that metal-electrorefining does not produce pure plutonium but transuranium elements, irradiation experiment of metal fuel with minor actinides is carried out by use of Phenix Fast Reactor in France. This article reports an overview of pyro-reprocessing and the present status of its research and development. The R and D activity proceeds to the process verification by use of genuine material and the development of engineering model of the process after finishing the verification of elemental technology. Irradiation study of metal fuel will be started by use of JOYO Fast Reactor as well as Phenix Fast Reactor. The target at 2015 is to finish the irradiation programs by both reactors and to demonstrate the pyro-process flow and related technologies by use of irradiated material. After finishing this stage, we expect to be technically feasible to design a pyro-process facility with a throughput of several tones of spent fuels. While R and D on pyro-technology has started initially in the U.S. and followed by CRIEPI, the several activities, currently, are followed in European and Asian nations. The engineering installation of electrochemical reduction successfully achieved by uranium test with 20 kg/batch and the construction of hot cell for handling a 20 kg/batch spent fuel finished in the Korean Atomic Energy Research Institute, KAERI. China has started R and D on metal fuel fast reactor and pyro-reprocessing as a

  7. The plutonium utilization in thermal and fast reactor in Japan

    Amanuma, T.; Uematsu, K.

    1977-01-01

    The nuclear power development in Japan is rather extensive one, and the installed nuclear power capacity is expected to reach 49,000 MWe by 1985 and possibly to reach 170,000 MWe by 2000 according to a prediction. Currently istalled nuclear power is mainly based on Light Water type Reactor, and this trend is expected to persist for the time-being. The plutonium produced by LWR will be accumulated to 20 tons by 1985 and to more than 200 tons by 2000. If the produced plutonium will simply be stored, it will raise the economic pressure to utilities and the management and physical protection problems associated with plutonium storing. Therefore, it is not too wise simply to store plutonium in a locked vault. In Japan, there are three ways of solving these problems which are currently worked out. There is no doubt that the best solution is to use plutonium in fast reactors. To reach this goal, an Experimental Fast Reactor ''JOYO''has been constructed and it is waiting for criticality in very near future. A prototype fast breeder reactor ''MONJU'', which is designed for about 300 MWe, is nearing to the last stage of the design work. The start of its construction will take place in a few yesars. The domonstration fast breeder reactor will come next to ''MONJU'' and the large scale commercial use of fast breeder reactor is expected to start around 1995. To anwer the near-term need for plutonium utilization, two technologies, which are equally important to Japan, are currently developed. One is the recycle use of plutonium into LWR. This technology has long been jointly developed by research organizations and utilities. Some of fuel irradiation data are already obtained and the physics study has also been extensive. The application of this technology is expected to start about 1987. The other is to burn plutonium in an Advanced Thermal Reactor (D 2 O moderated, Boiling Water Cooled) which shows better characteristics of using plutonium. The 160 MWe ''Fugen'' is a prototype

  8. Systemization of burnup sensitivity analysis code (2) (Contract research)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

  9. Real-time Environmental Monitoring Data on the internet

    Nakashima, N.I.; Maruo, Y.; Tobita, K.; Takeyasu, M.

    2000-01-01

    Japan Nuclear Cycle Development Institute (JNC) places great emphasis on safety, information disclosure and communication with the local community. The Real-time Environmental Monitoring Data (REMD) was made to provide to the public on the JNC web-site (http://www.jnc.go.jp/). It is the first organization having nuclear facilities in Japan to open REMD on the Internet web-site. JNC Tokai Works included Tokai Reprocessing Plant (TRP) started to open REMD in Oct. 1998. O-arai Engineering Center (OEC) included the Experimental Fast Reactor JOYO opened in April 1999. OEC produced this web-site in both Japanese and English (http://www.jnc.go.jp/zooarai/Oantai_e/html/index.html). REMD means airborne gamma radiation dose rate, and Meteorological Observation Data. Tokai Works has 13 Monitoring Posts/Stations and OEC has 8 Monitoring Posts to measure airborne gamma radiation dose. The data from these Monitoring Posts/Stations are shown on the web-site. The Meteorological Observation Data in this web-site are wind direction, wind speed, temperature, humidity, precipitation, and atmospheric stability. Atmospheric stability provides information on the state of the atmosphere concerned with air diffusion. REMD web-site provides all these data mentioned above as current data, data tables, trend graphs, and additional information. They are updated every hour. The current data are shown with a graphical map around the JNC site. Data tables are shown within 7 days. Daily highest and lowest temperature and precipitation are also shown as a table. There are three kinds of trend graphs of airborne radiation dose rate, the latest 24 hours trend graph, 48 hours, and 7 days. Each graph is shown with a graph of precipitation, so that variation of airborne gamma radiation with rainfall can be seen. Some explanations of this web-site are expressed as additional information. The topics of them are airborne Radiation, Meteorology, Radioactivity and Radiation, and Rainfall and Radiation. A set

  10. Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

    Furuta, H.; Tadokoro, H.; Imura, A.; Furuta, Y.; Suekane, F.

    2010-01-01

    Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

  11. Development of next generation code system as an engineering modeling language (5). Investigation on restructuring method of conventional codes into two-layer system

    Yokoyama, Kenji

    2006-10-01

    A proposed method for gradually restructuring to the two-level system of next generation analysis system by reusing the conventional analysis system, called 'incremental method', was applied and evaluated. The following functions were selected for the evaluation: Neutron diffusion calculation for the three-dimensional XYZ system based on finite differential method, and input utilities of the cross-section data file used in the conventional system. In order to evaluate the effect of the restructuring, 'Module Coupling Index (MCI)' and 'McCabe's Cyclomatic Complexity (MCC)' were used for quantifying the quality of the modular design and the complexity of the program sequence of each module. Although MCIs of each module before restructuring were mainly 6 - 7 degrees, it was possible to reduce them to under 4 degrees in most module by restructuring with the incremental method. And, it is found that the modules under 4 degrees of MCI can be easily combined with different programming languages, which are necessary for building the two-layer system. In the meantime, MCCs in most module before restructuring were over 20 and some were over 50. The incremental method could reduce them to under 10 when C++ was used, and reduce them to under 20 when FORTRAN was used. It is correspondent to reduction of the error frequency occurred in its modification from 20 - 40% to 5 - 10%. The total number of MCC could be reduced to 1/3 when C++ was used, and to 1/2 when FORTRAN was used. By using the restructured functions in the present study and some previously developed functions, a reactor analysis tool was systematized and applied to criticality analysis of the Experimental Fast Reactor 'JOYO' MR-I. In addition, the following two functionality expansion tests were performed: To add cross section direct perturbation functionality, and to add control rod criticality position search functionality. In the tests, both the functionality expansions were carried out satisfying the condition

  12. Sophistication of burnup analysis system for fast reactor (2)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  13. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  14. Electrochemical approach to corrosion behavior of ferritic steels in Flibe melt

    Nishimura, H.; Suzuki, A.; Terai, T.; Kondo, M.; Sagara, A.; Noda, N.

    2007-01-01

    Full text of publication follows: A mixture of LiF-BeF 2 , Flibe, is considered as a candidate material for tritium breeding in a fusion liquid blanket. Flibe has favorable characteristics such as high chemical stability and low electric conductivity. However, it produces TF with neutron irradiation, which is corrosive to structural materials. Therefore, the compatibility of structural materials with Flibe is a critical issue. Up to the present, the compatibility of some materials with Flibe was examined by carrying out simple immersion tests under limited conditions. By visual observations and analyses such as XRD on the surfaces after washing out Flibe from specimens, it was found that ferritic steels seemed to have good compatibility. However, strictly speaking, surface condition of the specimens should not be same as that during immersion in melt because these specimens were subjected to heat treatments and washing processes in order to remove solidified Flibe. Therefore, we planed electrochemical experiment to observe corrosion behavior during immersion. In this study, by carrying out cyclic voltammetry on specimens to observe alteration of surface condition of specimen in Flibe melt from moment to moment, the compatibility of ferritic steel with Flibe melt was discussed on. JLF-1 JOYO-II heat ferritic steel (Fe-9.000r-1.98W-0.09C-0.49Mn-0.20V-0.083Ta) which is a candidate low activation ferritic steel as a structural material of fusion reactor was chosen as a test specimen. Fe-9Cr and Fe-2W alloys were also chosen for comparison. The size of all specimens was 20 x 10 x 1 mm. A electrochemical cell was assembled using these specimens as working electrodes. Pt was chosen as a material for quasi-reference electrode. A Ni crucible which was the container of electrolyte, Flibe, was used as a counter electrode. 600 grams of Flibe was prepared and purified by HF/H 2 bubbling before being filled in the Ni crucible. Each specimen was dunked into Flibe at 773, 823 and

  15. Radiation control report on intermediate heat exchanger replacement and related works

    Kanou, Y.; Yamanaka, T.; Sasajima, T.; Hoshiba, H.; Emori, S.; Shindou, K.

    2002-03-01

    The 13th periodical inspection of the experimental fast reactor JOYO is being made from Jun. 2000 to Jan. 2003. While this inspection, from the end of Oct. 2000 to Nov. 2001, the MK-III modification work on heat transport system was made in lower region of the reactor containment vessel in the reactor facility (under floor area). In the MK-III modification work, the works important to radiation control were the replacement of intermediate heat exchangers (IHXs) and fixtures, and the picking out of the surveillance material from primary heat transport piping carried out in the maintenance building. Because the working areas of these works were executed in small space around the complicated primary heat transport piping, workability was bad and dose rate from the corrosion products (CP) in piping or fixtures was high. In such condition, radiation control was performed mainly concerned about external exposure. The planted total external exposure of the IHX replacement and related works was 7135 man-mSv (target of total dose control: less than 5708 man-mSv, 80% of the plan), derived from special radiation work plants for segmental works, concerned about work procedure, number of workers, period of work, dose rate of working area and surface dose rate of equipments. The special radiation control organization was established for such long and large-scale work. The spatial organization held detailed discussion about radiation control of this work with the execution section and contractors appropriately, performance careful external/internal exposure control and surface contamination control and made efforts to reduce te external exposure thoroughly. As a result of these action, the total external exposure was 2386 man·mSv (≅33% of the plan, ≅42% of the target) and the maximum individual exposure were 24.7 mSv for staffs and 21.7mSv for contractors. The dose rate, surface contamination and air contamination while the works were kept under the control level with the

  16. Adsorption properties of technetium and rhenium for hybrid microcapsules enclosing Toa extractant

    Wu, Y.; Mimura, H.; Niibori, Y. [Tohoku University, Graduate School of Engineering, Department of Quantum Science and Energy Engineering, Aramaki-Aza-Aoba 6-6-01-2, Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 Japan (Japan); Koyama, S.; Ohnishi, T., E-mail: yanwu@cyric.tohoku.ac.j [Japan Atomic Energy Agency, O-arai Research and Development Center, Fuels and Materials Department, Alpha-Gamma Section, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 Japan (Japan)

    2010-10-15

    Special attention has been given to the separation and recovery of the VII-group elements Tc and Re, in relation to the partitioning of high-level liquid waste (HLLW) generated from the nuclear fuel reprocessing process. In this study, a tertiary amine (tri-n-octylamine Toa), which is effective for the extraction of oxo anions, was encapsulated in a calcium alginate gel polymer (CaALG), and the adsorption behaviours of TcO{sub 4} and ReO{sub 4}{sup -} in the presence of nitric acid and hydrochloric acid were examined by using calcium alginate microcapsules (M Cs) enclosing Toa extractant (Toa-CaALG M Cs). The order of the distribution coefficient K{sub d} for different oxo anions at 0.1 M HNO{sub 3} was ReO{sub 4}{sup -}> WO{sub 4}{sup 2-}> CrO{sub 4}{sup 2-} {approx} MoO{sub 4}{sup 2-}>> SeO{sub 3}{sup 2-}. Toa-CaALG still exhibited high uptake ability for ReO{sub 4}{sup -} even after irradiation with {sup 60}Co {gamma}-rays (dose: 17.6 kGy). Uptake of TcO{sub 4}{sup -} in the presence of 1 M HNO{sub 3} was readily attained within 3 h. Relatively large K{sub d} values above 10{sup 2} cm{sup 3}/g were obtained for Toa-CaALG in the presence of 0.01 {approx} 1 M HNO{sub 3}. All of the TcO{sub 4}{sup -} was successfully adsorbed by Toa-CaALG from the simulated HLLW. The adsorbed TcO{sub 4}{sup -} was then effectively eluted with 5 M or 7 M HNO{sub 3} solution. Further, the selective uptake of ReO{sub 4}{sup -} (a chemical analogue of TcO{sub 4}{sup -}) was confirmed by using actual HLLW (Fbr, Joyo, JAEA), and uptake (%) above 99% was obtained. Toa-CaALG was thus effective for the selective separation and recovery of TcO{sub 4}{sup -} and ReO{sub 4}{sup -} from waste solutions containing highly concentrated HNO{sub 3} and NaNO{sub 3}. Microencapsulation techniques with alginate gel polymer can be applied to other ion exchangers and extractants, and the M Cs immobilizing these adsorbents are effective for the advanced separation of various radionuclides, rare metals and

  17. Study on an innovative fast reactor utilizing hydride neutron absorber - Final report of phase I study

    Konashi, K.; Iwasaki, T.; Itoh, K.; Hirai, M.; Sato, J.; Kurosaki, K.; Suzuki, A.; Matsumura, Y.; Abe, S.

    2010-01-01

    (Aluminum diffusion coating) and oxidation was developed and the hydrogen transfer coefficient was drastically reduced. As the hydride compatibility with sodium was confirmed by the experiment, a sodium bonding pin was also developed along with the helium bonding pin. In addition, the hydrides were irradiated in the experimental Fast Reactor 'Joyo' without any defects. The phase II study has started in 2009 to extend the research for the hydride absorber application to FBR. (authors)

  18. Systemization of burnup sensitivity analysis code. 2

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

  19. Corrosion of steels in molten gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li)

    Kondo, Masatoshi, E-mail: kondo.masatoshi@nr.titech.ac.jp [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Ishii, Masaomi [Department of Nuclear Engineering, School of Engineering, Tokai University, 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292 (Japan); Muroga, Takeo [Department of Helical Plasma Research, National Institute for Fusion Science, Toki, Gifu 502-5292 (Japan)

    2015-10-15

    Graphical abstract: Corrosion of RAFM steel, JLF-1, in liquid Sn–20Li was caused by the formation of Fe-Sn alloyed layer. - Highlights: • The corrosion tests were performed for the reduced activation ferritic martensitic steel JLF-1 and the austenitic steel SUS316 in liquid Ga, Sn and Sn-20Li at 873 K up to 750 h. • The weight loss of the specimens exposed to liquid Ga, Sn and Sn-20Li was evaluated. • The corrosion of the steels in liquid Ga was caused by the alloying reaction between Ga and Fe on the steel surface. • The corrosion of the steels in liquid Sn was caused by the alloying reaction between Sn and Fe on the steel surface. • The corrosion of the steels in liquid Sn-20Li was caused by the formation of the Fe-Sn alloyed layer and the diffusion of Sn and Li into the steel matrix. - Abstract: The compatibility of steels in liquid gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li) was investigated by means of static corrosion tests. The corrosion tests were performed for reduced activation ferritic martensitic steel JLF-1 (JOYO-HEAT, Fe–9Cr–2W–0.1C) and austenitic steel SUS316 (Fe–18Cr–12Ni–2Mo). The test temperature was 873 K, and the exposure time was 250 and 750 h. The corrosion of these steels in liquid Ga, Sn and Sn–20Li alloy was commonly caused by the formation of a reaction layer and the dissolution of the steel elements into the melts. The reaction layer formed in liquid Ga was identified as Fe{sub 3}Ga from the results of metallurgical analysis and the phase diagram. The growth rate of the reaction layer on the JLF-1 steel showed a parabolic rate law, and this trend indicated that the corrosion could be controlled by the diffusion process through the layer. The reaction layer formed in liquid Sn and Sn–20Li was identified as FeSn. The growth rate had a linear function with exposure time. The corrosion in Sn and Sn–20Li could be controlled by the interface reaction on the layer. The growth rate of the layer formed

  20. Corrosion of steels in molten gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li)

    Kondo, Masatoshi; Ishii, Masaomi; Muroga, Takeo

    2015-01-01

    Graphical abstract: Corrosion of RAFM steel, JLF-1, in liquid Sn–20Li was caused by the formation of Fe-Sn alloyed layer. - Highlights: • The corrosion tests were performed for the reduced activation ferritic martensitic steel JLF-1 and the austenitic steel SUS316 in liquid Ga, Sn and Sn-20Li at 873 K up to 750 h. • The weight loss of the specimens exposed to liquid Ga, Sn and Sn-20Li was evaluated. • The corrosion of the steels in liquid Ga was caused by the alloying reaction between Ga and Fe on the steel surface. • The corrosion of the steels in liquid Sn was caused by the alloying reaction between Sn and Fe on the steel surface. • The corrosion of the steels in liquid Sn-20Li was caused by the formation of the Fe-Sn alloyed layer and the diffusion of Sn and Li into the steel matrix. - Abstract: The compatibility of steels in liquid gallium (Ga), tin (Sn) and tin lithium alloy (Sn–20Li) was investigated by means of static corrosion tests. The corrosion tests were performed for reduced activation ferritic martensitic steel JLF-1 (JOYO-HEAT, Fe–9Cr–2W–0.1C) and austenitic steel SUS316 (Fe–18Cr–12Ni–2Mo). The test temperature was 873 K, and the exposure time was 250 and 750 h. The corrosion of these steels in liquid Ga, Sn and Sn–20Li alloy was commonly caused by the formation of a reaction layer and the dissolution of the steel elements into the melts. The reaction layer formed in liquid Ga was identified as Fe 3 Ga from the results of metallurgical analysis and the phase diagram. The growth rate of the reaction layer on the JLF-1 steel showed a parabolic rate law, and this trend indicated that the corrosion could be controlled by the diffusion process through the layer. The reaction layer formed in liquid Sn and Sn–20Li was identified as FeSn. The growth rate had a linear function with exposure time. The corrosion in Sn and Sn–20Li could be controlled by the interface reaction on the layer. The growth rate of the layer formed in

  1. Reactor Dosimetry State of the Art 2008

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

    2009-08-01

    data, damage correlations. Two-dimensional mapping of the calculated fission power for the full-size fuel plate experiment irradiated in the advanced test reactor / G. S. Chang and M. A. Lillo. The radiation safety information computational center: a resource for reactor dosimetry software and nuclear data / B. L. Kirk. Irradiated xenon isotopic ratio measurement for failed fuel detection and location in fast reactor / C. Ito, T. Iguchi and H. Harano. Characterization of dosimetry of the BMRR horizontal thimble tubes and broad beam facility / J.-P. Hu, R. N. Reciniello and N. E. Holden. 2007 nuclear data review / N. E. Holden. Further dosimetry studies at the Rhode Island nuclear science / R. N. Reciniello ... [et al.]. Characterization of neutron fields in the experimental fast reactor Joyo MK-III core / S. Maeda ... [et al.]. Measuring [symbol]Li(n, t) and [symbol]B(n, [symbol]) cross sections using the NIST alpha-gamma apparatus / M. S. Dewey ... [et al.]. Improvement of neutron/gamma field evaluation for restart of JMTR / Y. Nagao ... [et al.]. Monitoring of the irradiated neutron fluence in the neutron transmutation doping process of HANARO / M.-S. Kim and S.-J. Park.Training reactor VR-l neutron spectrum determination / M. Vins, A. Kolros and K. Katovsky. Differential cross sections for gamma-ray production by 14 MeV neutrons on iron and bismuth / V. M. Bondar ... [et al.]. The measurements of the differential elastic neutron cross-sections of carbon for energies from 2 to 133 ke V / O. Gritzay ... [et al.]. Determination of neutron spectrum by the dosimetry foil method up to 35 Me V / S. P. Simakov ... [et al.]. Extension of the BGL broad group cross section library / D. Kirilova, S. Belousov and Kr. Ilieva. Measurements of neutron capture cross-section for tantalum at the neutron filtered beams / O. Gritzayand V. Libman. Measurements of microscopic data at GELINA in support of dosimetry / S. Kopecky ... [et al.]. Nuclide guide and international chart of

  2. Establishing a Scientific Basis for Optimizing Compositions, Process Paths and Fabrication Methods for Nanostructured Ferritic Alloys for Use in Advanced Fission Energy Systems

    Odette, G Robert; Cunningham, Nicholas J., Wu, Yuan; Etienne, Auriane; Stergar, Erich; Yamamoto, Takuya

    2012-02-21

    lowest Y2O3 concentration of 0.2 wt.%. An APT characterization of MA957 joined by friction stir welding (FSW) showed that this solid sate joining procedure had only a modest effect on the NF number density (N) and average diameter () compared to an as extruded sample. FSW appears to rearrange the NFs, which become highly aligned with sub-boundary and dislocation structures to an extent that are not observed in the as extruded case. The aligned NF structures are less apparent, but seem to persist after post weld annealing at 1150ºC for 3 h following which reduces N, consistent with a significant reduction in hardness. Lastly, several NFA materials, including MA957 and various 14YWT alloys, have been included in irradiation experiments performed at the Advanced Test Reactor, the JOYO sodium cooled fast reactor, the High Flux Isotope Reactor, and the SINQ spallation neut

  3. Back-to-back technical meetings (TMs): 'TM on the coordinated project (CRP) analyses of and lessons learned from the operational experience with fast reactor equipment and systems' and 'TM to coordinate the Agency's fast reactor knowledge preservation international project in Russia'. Working material

    NONE

    2005-07-01

    Since the early 1960's, several countries have undertaken important fast breeder reactor development programs. Fast test reactors were constructed and successfully operated in a number of countries, including Rapsodie (France), KNK-II (Germany), FBTR (India), JOYO (Japan), DFR (UK), BR-10, BOR-60 (Russia), and EBR-II, Fermi, FFTF (USA). This was followed by commercial size prototypes (Phenix, Superphenix (France), SNR-300 (Germany), MONJU (Japan), PFR (UK), BN-350 (Kazakhstan), BN-600 (Russia)], either just under construction, coming on line, or experiencing long term operation. However, from the 1980s onward, and mostly for economical and political reasons, fast reactor development in general began to decline. By 1994, in the USA, the Clinch River Breeder Reactor (CRBR) had been cancelled, and the two fast reactor test facilities, FFTF and EBR-II had been shutdown - EBRII permanently, and FFTF, until recently, in standby condition, but now also facing permanent closure. Thus, in the U.S., effort essentially disappeared for fast breeder reactor development. Similarly, programs in other nations were terminated or substantially reduced. In France, Superphenix was shut down at the end of 1998; SNR-300 in Germany was completed but not taken into operation, and KNK-II was permanently shut down in 1991 (after 17 years of operation) and is scheduled to be dismantled by 2004. In the UK, PFR was shut down in 1994, and in Kazakhstan, BN-350 was shut down in 1998. As the interest and activity in the fast breeder reactor diminished, the retirement of many of the developers and acknowledged experts of this technology reached its peak, between 1990 and 2000. The effort and investment required to replace these skills also diminished in parallel. In addition, the facilities (e.g., hot cells, fuel fabrication and inspection lines, seismic test rigs) required to develop and maintain the fast reactor program are drifting into a degraded state or are being shut down. This leads to the