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Sample records for reactor joyo level

  1. Introduction of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  3. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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)

  6. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  8. Radiation protection monitoring at the JOYO experimental fast reactor

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  9. Joyo progress report, vol. 8

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  11. 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

    International Nuclear Information System (INIS)

    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)

  12. JOYO construction and preoperational test experience

    International Nuclear Information System (INIS)

    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.)

  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

    International Nuclear Information System (INIS)

    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. 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

    International Nuclear Information System (INIS)

    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.)

  15. 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

    International Nuclear Information System (INIS)

    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.)

  16. Power coefficient anomaly in JOYO

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Yoshimi, H.; Hachiya, Y.

    1990-01-01

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

  18. 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

    International Nuclear Information System (INIS)

    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.)

  19. Improvement of seismic observation systems in JOYO

    International Nuclear Information System (INIS)

    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.)

  20. The JOYO remote monitoring system

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  2. 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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  6. 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

    International Nuclear Information System (INIS)

    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)

  7. Operation experiences of JOYO fuel failure detection system

    International Nuclear Information System (INIS)

    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)

  8. Power coefficient anomaly in Joyo, (2)

    International Nuclear Information System (INIS)

    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.)

  9. JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    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)

  10. Joyo ATWS test analysis by Mimir-N2

    International Nuclear Information System (INIS)

    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)

  11. Development of the flow control irradiation facility for JOYO

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    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)

  15. 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'

    International Nuclear Information System (INIS)

    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)

  16. Reactor water level control device

    International Nuclear Information System (INIS)

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  2. JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  6. Operational experience of the fuel cleaning facility of Joyo

    International Nuclear Information System (INIS)

    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)

  7. Operational experience of the fuel cleaning facility of Joyo

    Energy Technology Data Exchange (ETDEWEB)

    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)

  8. Development of JOYO operational guidance system for emergency condition

    International Nuclear Information System (INIS)

    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)

  9. Processing of Refractory Metal Alloys for JOYO Irradiations

    International Nuclear Information System (INIS)

    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. Development of JOYO plant operation management expert tool

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    International Nuclear Information System (INIS)

    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)

  12. Reactor Emergency Action Level Monitor

    International Nuclear Information System (INIS)

    Touchton, R.A.

    1987-01-01

    The Reactor Emergency Action Level Monitor (REALM) Expert System is designed to provide assistance in the identification of a nuclear power plant emergency situation and the determination of its severity. REALM has been developed to operate in a real-time processing environment. REALM embodies a hybrid architecture utilizing both rule-based reasoning and object-oriented programming techniques borrowed from the Artificial Intelligence discipline of Computer Sciences. The rulebase consists of event-based rules and symptom-based rules. The symptom-based rules go beyond the current EAL structure to address the more problematic scenarios and entail a more symbolic representation of the plant information. The results to date have been encouraging that expert system technology can provide improved emergency decision-making capability in nuclear power plants

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

    International Nuclear Information System (INIS)

    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. Reactor water level measuring device

    International Nuclear Information System (INIS)

    Kuroki, Reiji; Asano, Tamotsu.

    1996-01-01

    A condensation vessel is connected to the upper portion of a reactor pressure vessel by way of a pipeline. The lower portion of the condensation vessel is connected to a low pressure side of a differential pressure transmission device by way of a reference leg pipeline. The high pressure side of the differential pressure transmission device is connected to the lower portion of the pressure vessel by way of a pipeline. The condensation vessel is equipped with a temperature sensor. When a temperature of a gas phase portion in the condensation vessel is lowered below a predetermined level, and incondensible gases in the condensation vessel starts to be dissolved in water, signals are sent from the temperature sensor to a control device and a control valve is opened. With such a constitution, CRD driving water flows into the condensation vessel, and water in which gases at the upper portion of the condensation vessel is dissolved flows into the pressure vessel by way of a pipeline. Then, gases dissolved in a reference water column in the reference leg pipeline are eliminated and the value of a reference water pressure does not change even upon abrupt lowering of pressure. (I.N.)

  15. Development of JOYO MK-II core characteristics database

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  17. Water level monitoring device in nuclear reactor

    International Nuclear Information System (INIS)

    Miura, Kiyohide; Otake, Tomohiro.

    1988-01-01

    Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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)

  1. Closeout of JOYO-1 Specimen Fabrication Efforts

    International Nuclear Information System (INIS)

    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

  2. Method of measuring reactor water level

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1979-01-01

    Purpose: To provide a water level measuring system so that a reactor water level detecting signal can be corrected in correspondence to a recirculation flow, thereby to carry out a correct water level detection in a wide range of the reactor. Method: According to the operation record of a precursor reactor, the ratio Δh of the lowering of the water level due to the recirculation flow is lowered in proportion to the ratiowith respect to the rated differential pressure of the recirculation flow. Accordingly, the flow of recirculation pump is measured by an elbow differential pressure generator utilizing an elbow of a pipe, and the measured value is multiplied by a gain by a ratio setter, and therefter, an addition computation is carried out by an adder for correcting the signal from a water level detector. When the signal from the water level detector is corrected in this manner, the influence of the lowering of the water level due to the recirculation flow can be removed, and an interlocker predetermined in the defined water level can be actuated, thus the influence of the dynamic pressure due to the recirculation flow acting on the instrumental pipe line detecting the reactor water level can be removed effectively. (Yoshino, Y.)

  3. PSA Level 2 activities for RBMK reactors

    International Nuclear Information System (INIS)

    Gubler, R.

    1998-01-01

    Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

  4. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  9. Level tracking in detailed reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Aktas, B.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

    We introduce a useful test problem for judging the performance of reactor safety codes in situations where moving two-phase mixture levels are present. The test problem tracks a two-phase liquid level as it rises and then falls back to its original position. Pure air exists above the level, and a low void air-water mixture is below the level. Conditions are subcooled and isothermal to remove complications resulting from failures of interfacial heat transfer packages to properly account for the level. Comparisons are made between the performance of current versions of CATHARE, RELAP5, TRAC-BF1, and TRAC-PF1. These system codes are based on finite-difference methods with a fixed, Eulerian staggered grid in space. When a partially filled cell with a mixture level discontinuity becomes the donor cell, the sharp changes in fluid properties across the interface results in numerical oscillations of various terms. Furthermore, the cell-to-cell convection of mass, momentum and energy are inaccurately predicted nearby a mixture level. To adequately model moving mixture levels, an efficient discontinuity tracking method for the finite-difference Eulerian approximations is described. This model had been implemented in the TRAC-BWR code for the two-phase mixture level tracking since the TRAC-BD1 Version (released April 1984). The result of the test problem run by the current version of TRAC-BF1/MOD1 with the mixture level tracking model shows some peculiar behavior of the variables such as velocities, pressures and interfacial terms. A systematic approach to improving performance of the tracking method is described. Implementing this approach in TRAC-BF1/MOD1 has shown a major improvement in the results.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  12. Ultrasonic level and temperature sensor for power reactor applications

    International Nuclear Information System (INIS)

    Dress, W.B.; Miller, G.N.

    1983-01-01

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel

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

    International Nuclear Information System (INIS)

    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)

  14. Suppression device for the reactor water level lowering

    International Nuclear Information System (INIS)

    Kasuga, Hajime; Kasuga, Hiroshi.

    1984-01-01

    Purpose: To suppress the lowering in the reactor water level so as to avoid unnecessary actuation of ECCS upon generation of transient changes which forecasts the lowering of the reactor water level in a BWR type reactor. Constitution: There are provided a water level suppression signal generator for generating a water level suppression signal upon generation of a transient change signal which forecasts the water level lowering in a nuclear reactor and a recycling flow rate controller that applies a recycling flow rate control signal to a recycling pump drive motor by the water level lowering suppression signal. The velocity of the recycling pump is controlled by a reactor scram signal by way of the water level lowering suppresion signal generator and a recycling flow rate controller. Then, the recycling reactor core flow rate is decreased and the void amount in the reactor is transiently increased where the water level tends to increase. Accordingly, the water level lowering by the scram is moderated by the increasing tendency of the water level. (Ikeda, J.)

  15. An appropriate level for reactor regulatory research

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Kouts, H.J.C.

    1986-01-01

    In this brief paper the appropriate role for NRC's research program is set down, based on the experience of the last decade; and the broad outlines of an appropriate research program for the next decade are discussed. The two authors bring to this topic direct personal experience: they are two of only three former directors of NRC's research program in its history (the other was the last Saul Levine). The discussion considers only reactor safety research, although NRC's research program covers several other areas as well. Safeguards, environmental and health studies, waste management, and transportation are not covered

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

    International Nuclear Information System (INIS)

    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

  17. A Review of the Fast Reactor Programme in Japan

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. Effect of reactor conditions on MSIV-ATWS power level

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1987-01-01

    In a boiling water reactor (BWR) when there is closure of the main steam isolation valves (MSIVs), the energy generated in the core will be transferred to the pressure suppression pool (PSP) via steam that flows out of the relief valves. The pool has limited capacity as a heat sink and hence, if there is no reactor trip [an anticipated transient without scram (ATWS) event], there is the possibility that the pool temperature may rise beyond acceptable limits. The present study was undertaken to determine how the initial reactor conditions affect the power level during an MSIV-ATWS event. The time of interest is the 20- to 30-min period when it is assumed that the reactor is in a quasi equilibrium condition with the water level and pressure fixed, natural circulation conditions and no control rod movement or significant boron in the core. The initial conditions of interest are the time of the cycle and the operating state

  20. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  1. Monitoring PWR reactor vessel liquid level with SPNDs during LOCAs

    International Nuclear Information System (INIS)

    Adams, J.P.

    1982-01-01

    Data from in-core self-powered neutron detectors taken during two nuclear loss-of-coolant accident simulations have been correlated with core moderator density changes. The detector current attenuation has been calculated during blowdown and reflood phases of the simulation. Based on these data, it is concluded that these detectors could be used to monitor reactor vessel liquid level during loss-of-coolant accidents in pressurized water reactors

  2. Design of megawatt power level heat pipe reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-12

    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

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

    International Nuclear Information System (INIS)

    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)

  4. Calculations of radiation levels during reactor operations for safeguard inspections

    International Nuclear Information System (INIS)

    Sobhy, M.

    1996-01-01

    When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity. (author)

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

    International Nuclear Information System (INIS)

    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.)

  6. Evaluation of the activity levels in fusion reactor blankets

    International Nuclear Information System (INIS)

    Gruber, J.

    1977-05-01

    The activation of a fusion reactor blanket (316 SS or V-10Cr-10Ti as structure) with a minimum lithium inventory has been calculated for 0.83 MW/m 2 wall load. The resulting radiation levels and waste problems are discussed. The dose rate near the steel structure will always be higher than 0.1 rem/h due to its niobium content. After 200 to 100,000 years of decay the potential biological hazard originating from this high level fusion reactor waste (with plutonium recyclation). (orig.) [de

  7. quantification of emergency action levels for research reactor

    International Nuclear Information System (INIS)

    Wu Zhongwang; Qu Jingyuan; Liu Yuanzhong; Xi Shuren

    2000-01-01

    Emergency action level (EAL) technical criteria or parameters for emergency conditions classes. Reference methodology for development of EAL in foreign countries, in process of developed and reviewed emergency plan of home several research reactors, the author thought that should be taken initiating conditions which result in emergency conditions quantified some instrumental readings or alarm thresholds, in order to distinguish and confirm emergency conditions and provide technical bases for emergency response actions. Then based on this principle, revised or developed emergency plans of INET Tsinghua University, promote development of work for emergency plan of research reactors

  8. Severe accident analysis for level 2 PSA of SMART reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

    2010-12-15

    The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

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

    International Nuclear Information System (INIS)

    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)

  10. Development of reactor water level sensor for extreme conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miura, K; Ogasawara, T [Sukegawa Electric Co., Ltd., Hitachi, Ibaraki (Japan); Shibata, Akira; Nakamura, Jinichi; Saito, Takashi; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    In the Fukushima accident, measurement failure of water level was one of the most important factors which caused serious situation. The differential pressure type water level indicators are widely used in various place of nuclear power plant but after the accident of TMI-2, the need of other reliable method has been required. The BICOTH type and the TRICOTH type water level indicator for light water power reactors had been developed for in-pile water level indicator but currently those are not adopted to nuclear power plant. In this study, the development of new type water level indicator composed of thermocouple and heater is described. Demonstration test and characteristic evaluation of the water level indicator were performed and we had obtained satisfactory results. (author)

  11. Transient response of level instruments in a research reactor

    International Nuclear Information System (INIS)

    Cheng, Lap Y.

    1989-01-01

    A numerical model has been developed to simulate the dynamics of water level instruments in a research nuclear reactor. A bubble device, with helium gas as the working fluid, is used to monitor liquid level by sensing the static head pressure due to the height of liquid in the reactor vessel. A finite-difference model is constructed to study the transient response of the water level instruments to pressure perturbations. The field equations which describe the hydraulics of the helium gas in the bubbler device are arranged in the form of a tridiagonal matrix and the field variables are solved at each time step by the Thomas algorithm. Simulation results indicate that the dynamic response of the helium gas depends mainly on the volume and the inertia of the gas in the level instrument tubings. The anomalies in the simulated level indication are attributed to the inherent lag in the level instrument due to the hydraulics of the system. 1 ref., 5 figs

  12. Derivation methods for clearance levels applied to reactors

    International Nuclear Information System (INIS)

    Okoshi, Minoru; Seki, Takeo

    2001-01-01

    In order to support the discussion by the Nuclear Safety Commission, JAERI derived the unconditional clearance levels for concrete and metal arising from the operation and dismantling of nuclear reactors. The clearance levels of 20 radionuclides were derived from 10 μSv/y of individual doses by deterministic approach. In this approach, calculation models were established to assess individual doses resulting from 73 exposure pathways related to disposal and recycle/reuse, and realistic parameter values were selected considering Japanese natural and social conditions. The appropriateness of selected parameter values was confirmed by stochastic analyses. (author)

  13. New reactor safety circuit for low-power-level operation

    International Nuclear Information System (INIS)

    McDowell, W.P.; Keefe, D.J.; Rusch, G.K.

    1978-01-01

    In the operation of nuclear reactors at low-power levels, one of the primary instrumentation problems is that the statistical fluctuations of reactor neutron population are accentuated by conventional log-count-rate and differentiating circuits and can cause frequent spurious scrams unless long time constants are incorporated in the circuit. Excessive time constants may introduce undesirable delay in the circuit response to legitimate scram signals. The paper develops the concept of a count doubling-time monitor which generates a scram signal if the number of counts from a pulse type neutron detector doubles in a given period of time. The paper demonstrates the theoretical relation between count doubling time and asymptomatic periods. A practical circuit to implement the function is described

  14. Elementary migration around the Oklo nuclear reactors. Implications for high level radioactive wastes storage

    International Nuclear Information System (INIS)

    Menet-Dressayre, C.; Menager, M.T.

    1993-01-01

    The study of Uranium and rare earths near the reactors has displayed the radioelements transfer in the reactors neighbourhood. The main implications for high level radioactive wastes disposal in geological formations are discussed. 12 refs

  15. Understanding to requirements for educational level in qualification of reactor operators

    International Nuclear Information System (INIS)

    Zhang Chi; Yang Di; Zhou Limin

    2007-01-01

    Requirements for qualification of reactor operators in nuclear safety regulations were discussed in this paper. The new issue was described in the confirmation of education level of reactor operators. The understanding to the requirements for Educational Level in Qualification of Reactor Operators was provided according to Higher Education Law of the People's Republic of China. It was proposed to improve the confirmation of qualification of reactor operators as soon as possible. (authors)

  16. Reactor Emergency Action Level Monitor: Knowledge acquisition experiences

    International Nuclear Information System (INIS)

    Touchton, R.A.

    1987-01-01

    This paper presents the Knowledge Acquisition experiences in developing the Reactor Emergency Action Level Monitor (REALM) Expert System Prototype. REALM is an expert system which interprets plant sensor data and provides advice on the proper emergency classification. The REALM project is being funded by the Electric Power Research Institute, Consolidated Edison is serving as the host utility, and the effort is being conducted by Technology Applications, Inc. REALM is being designed to provide expert assistance in the identification of a nuclear power plant emergency situation and the determination of its severity, ultimately operating in a real-time, on-line processing environment. The paper discusses briefly the direct knowledge acquisition techniques used by the project team (who are themselves power industry engineers), to extract relevant knowledge from plant specifications and procedures

  17. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    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)

  18. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  19. The development of breeder reactors in Japan

    International Nuclear Information System (INIS)

    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

  20. Operational intervention levels for reactor emergencies IAEA recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Aguiar, Laís A. de; Reis, Arlene A. dos; Santos, Raul dos, E-mail: laguiar@ird.gov.br, E-mail: arlene@ird.gov.br, E-mail: raul@ird.gov.br [Instituto de Radioproteção e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The IAEA publication EPR-NPP-OILs-2017, Operational Intervention Levels for Reactor Emergences and Methodology for their Derivation, provides selected default OIL values, describing a methodology for their derivation, as well as practical tools and recommendations for their use. IAEA recommends that tools and default OIL values be directly integrated into national emergency arrangements or reviewed and modified as necessary to meet the specific emergency preparedness and response arrangements. The Institute of Radiation Protection and Dosimetry (IRD) has a Radiological Assessment TEAM (EAR) as pare of its Radiation Emergency Response System. Brazilian regulatory standards address actions for radiation emergencies encompassing necessary measures to assess public exposures, intervention levels to protect the public and recommendations for protective actions as, evacuation, relocation, sheltering and food restrictions. The objective of this paper is to present a discussion the use these OILs, to compare those ones established by the Brazilian standards and to propose a methodology on how OILs can be used by EAR/IRD in case of an emergency at the Brazilian NPP. (author)

  1. Operational intervention levels for reactor emergencies IAEA recommendations

    International Nuclear Information System (INIS)

    Aguiar, Laís A. de; Reis, Arlene A. dos; Santos, Raul dos

    2017-01-01

    The IAEA publication EPR-NPP-OILs-2017, Operational Intervention Levels for Reactor Emergences and Methodology for their Derivation, provides selected default OIL values, describing a methodology for their derivation, as well as practical tools and recommendations for their use. IAEA recommends that tools and default OIL values be directly integrated into national emergency arrangements or reviewed and modified as necessary to meet the specific emergency preparedness and response arrangements. The Institute of Radiation Protection and Dosimetry (IRD) has a Radiological Assessment TEAM (EAR) as pare of its Radiation Emergency Response System. Brazilian regulatory standards address actions for radiation emergencies encompassing necessary measures to assess public exposures, intervention levels to protect the public and recommendations for protective actions as, evacuation, relocation, sheltering and food restrictions. The objective of this paper is to present a discussion the use these OILs, to compare those ones established by the Brazilian standards and to propose a methodology on how OILs can be used by EAR/IRD in case of an emergency at the Brazilian NPP. (author)

  2. Low and intermediate level radioactive waste processing in plasma reactor

    International Nuclear Information System (INIS)

    Sauchyn, V.; Khvedchyn, I.; Van Oost, G.

    2013-01-01

    Methods of low and intermediate level radioactive waste processing comprise: cementation, bituminization, curing in polymer matrices, combustion and pyrolysis. All these methods are limited in their application in the field of chemical, morphological, and aggregate composition of material to be processed. The thermal plasma method is one of the universal methods of RAW processing. The use of electric-arc plasma with mean temperatures 2000 - 8000 K can effectively carry out the destruction of organic compounds into atoms and ions with very high speeds and high degree of conversion. Destruction of complex substances without oxygen leads to a decrease of the volume of exhaust gases and dimension of gas cleaning system. This paper presents the plasma reactor for thermal processing of low and intermediate level radioactive waste of mixed morphology. The equipment realizes plasma-pyrolytic conversion of wastes and results in a conditioned product in a single stage. As a result, the volume of conditioned waste is significantly reduced (more than 10 times). Waste is converted into an environmentally friendly form that suits long-term storage. The leaching rate of macro-components from the vitrified compound is less than 1.10 -7 g/(cm 2 .day). (authors)

  3. Immobilisation of high level nuclear reactor wastes in SYNROC

    Energy Technology Data Exchange (ETDEWEB)

    Ringwood, A E; Kesson, S E; Ware, N G; Hibberson, W; Major, A [Australian National Univ., Canberra. Inst. of Advanced Studies

    1979-03-15

    It is stated that the elements occurring in high-level nuclear reactor wastes can be safely immobilised by incorporating them within the crystal lattices of the constituent minerals of a synthetic rock (SYNROC). The preferred form of SYNROC can accept up to 20% of high level waste calcine to form dilute solid solutions. The constituent minerals, or close structural analogues, have survived in a wide range of geochemical environments for periods of 20 to 2,000 Myr whilst immobilising the same elements present in nuclear wastes. SYNROC is unaffected by leaching for 24 hours in pure water or 10 wt % NaCl solution at high temperatures and pressure whereas borosilicate glasses completely decompose in a few hours in much less severe hydrothermal conditions. The combination of these leaching results with the geological evidence of long-term stability indicates that SYNROC would be vastly superior to glass in its capacity to safely immobilise nuclear wastes, when buried in a suitable geological repository. A dense, compact, mechanically strong form of SYNROC suitable for geological disposal can be produced by a process as economical as that which incorporates radioactive waste in borosilicate glasses.

  4. Level 3 decommissioning of Triton - Nereide research reactor

    International Nuclear Information System (INIS)

    Lopes, E.; Pillette-Cousin, L.

    2002-01-01

    The French Atomic Energy Commission Center located at Fontenay-Aux-Roses has launched an extensive programme of site cleanup and decommissioning of nuclear facilities. This programme includes the level 3 decommissioning of the Triton and Nereide piles. These pool type research reactors were constructed in the late 1950's, primarily for R and D activities related to neutron physics studies, radiological shielding experiments and radioelement production. As of 1982, a level 2 decommissioning was achieved and over the the last twenty years, no activities were carried out in the facility. During 2001, there has been extensive investigation work carried out to acquire a better knowledge of the radiological status of the facility, in order to set up dismantling scenarios and to reduce the volume of generated radioactive waste. Indeed, one of the first and main operations to be carried out for dismantling Triton and Nereide piles is waste zoning, by using the facility layout, operating conditions and history, as well as the present radiological inventory. The paper describes the investigations and studies carried out to implement waste zoning. The paper also describes the preliminary dismantling operations undertaken on equipment and studies conducted to optimize the dismantling and cleanup of the facility. Finally, the paper presents the outline of the preferred dismantling and decommissioning options and the progress of the work to date. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  7. Application of fuzzy synthetic assessment to assess human factors design level on reactor control panel

    International Nuclear Information System (INIS)

    Peng Xuecheng

    1999-01-01

    Reactor control panel design level on human factors must be considered by designer. The author evaluated the human factor design level of arrangement and combinations including the switch buttons, meter dials and indication lamps on Minjiang Reactor and High-Flux Engineer Test Reactor (HFETR) critical device by application of fuzzy synthetic assessment method in mathematics. From the assessment results, the advantages and shortcomings are fount, and some modification suggestions have also been proposed

  8. Evaluation of strategies for end storage of high-level reactor fuel

    International Nuclear Information System (INIS)

    2001-01-01

    This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized

  9. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    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. Time to reach a given level of number of neutrons is stochastic analog of reactor period

    International Nuclear Information System (INIS)

    Ryazanov, V.V.

    2012-01-01

    In theory and in practice the operation of nuclear reactors to control the safety of the reactor is widely used deterministic value - the period of the reactor. It is proposed along with the period of the reactor using a stochastic analogue of this magnitude - a random amount of time to achieve a given level of a random process for the number of neutrons in the reactor. The paper discusses various features of the behavior of the mean and variance of time to achieve a specified level. This kind of features can be associated with impaired behavior of the reactor system. Introduced the value of time required to reach the level can be used to monitor and improve the safety of nuclear power plants

  11. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  12. Swelling, mechanical properties, and microstructure of Type 316 stainless steel at fusion reactor damage levels

    International Nuclear Information System (INIS)

    Horak, J.A.; Bloom, E.E.; Grossbeck, M.L.; Maziasz, P.J.; Stiegler, J.O.; Wiffen, F.W.

    1979-01-01

    Alloys such as AISI 316 stainless steel exhibit more swelling and larger decreases in ductility when irradiated to produce fusion reactor He and dpa levels than at fast reactor He and dpa levels. For T approx. 0 C to ensure adequate ductility for long-term service

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  16. A Single-Granule-Level Approach Reveals Ecological Heterogeneity in an Upflow Anaerobic Sludge Blanket Reactor.

    Directory of Open Access Journals (Sweden)

    Kyohei Kuroda

    Full Text Available Upflow anaerobic sludge blanket (UASB reactor has served as an effective process to treat industrial wastewater such as purified terephthalic acid (PTA wastewater. For optimal UASB performance, balanced ecological interactions between syntrophs, methanogens, and fermenters are critical. However, much of the interactions remain unclear because UASB have been studied at a "macro"-level perspective of the reactor ecosystem. In reality, such reactors are composed of a suite of granules, each forming individual micro-ecosystems treating wastewater. Thus, typical approaches may be oversimplifying the complexity of the microbial ecology and granular development. To identify critical microbial interactions at both macro- and micro- level ecosystem ecology, we perform community and network analyses on 300 PTA-degrading granules from a lab-scale UASB reactor and two full-scale reactors. Based on MiSeq-based 16S rRNA gene sequencing of individual granules, different granule-types co-exist in both full-scale reactors regardless of granule size and reactor sampling depth, suggesting that distinct microbial interactions occur in different granules throughout the reactor. In addition, we identify novel networks of syntrophic metabolic interactions in different granules, perhaps caused by distinct thermodynamic conditions. Moreover, unseen methanogenic relationships (e.g. "Candidatus Aminicenantes" and Methanosaeta are observed in UASB reactors. In total, we discover unexpected microbial interactions in granular micro-ecosystems supporting UASB ecology and treatment through a unique single-granule level approach.

  17. A Single-Granule-Level Approach Reveals Ecological Heterogeneity in an Upflow Anaerobic Sludge Blanket Reactor

    Science.gov (United States)

    Mei, Ran; Narihiro, Takashi; Bocher, Benjamin T. W.; Yamaguchi, Takashi; Liu, Wen-Tso

    2016-01-01

    Upflow anaerobic sludge blanket (UASB) reactor has served as an effective process to treat industrial wastewater such as purified terephthalic acid (PTA) wastewater. For optimal UASB performance, balanced ecological interactions between syntrophs, methanogens, and fermenters are critical. However, much of the interactions remain unclear because UASB have been studied at a “macro”-level perspective of the reactor ecosystem. In reality, such reactors are composed of a suite of granules, each forming individual micro-ecosystems treating wastewater. Thus, typical approaches may be oversimplifying the complexity of the microbial ecology and granular development. To identify critical microbial interactions at both macro- and micro- level ecosystem ecology, we perform community and network analyses on 300 PTA–degrading granules from a lab-scale UASB reactor and two full-scale reactors. Based on MiSeq-based 16S rRNA gene sequencing of individual granules, different granule-types co-exist in both full-scale reactors regardless of granule size and reactor sampling depth, suggesting that distinct microbial interactions occur in different granules throughout the reactor. In addition, we identify novel networks of syntrophic metabolic interactions in different granules, perhaps caused by distinct thermodynamic conditions. Moreover, unseen methanogenic relationships (e.g. “Candidatus Aminicenantes” and Methanosaeta) are observed in UASB reactors. In total, we discover unexpected microbial interactions in granular micro-ecosystems supporting UASB ecology and treatment through a unique single-granule level approach. PMID:27936088

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

    International Nuclear Information System (INIS)

    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)

  19. Determination of the lowest critical power levels of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Binh, Do Quang; Nghiem, Huynh Ton; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    This paper presents the experimental methods for determining critical states of the Dalat Nuclear Research Reactor containing an extraneous neutron source induced by gamma ray reactions on beryllium in the reactor. The lowest critical power levels are measured at various moments after the reactor is shut down following 100 hours of its continuous operation. Th power levels vary from (0.5-1.2) x 10{sup -4} of P{sub n}, i.e. (25-60)W to (1.1-1.6) x 10{sup -5} of P{sub n}, i.e. (5.5-8)W at corresponding times of 4 days to 13 days after the reactor is shut down. However the critical power must be chosen greater than 500 W to sustain the steady criticality of the reactor for a long time. (author). 3 refs. 4 figs. 1 tab.

  20. A review of fast reactor progress in Japan

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Upgrade of reactor operation technology

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Malathi, N.; Sahoo, P., E-mail: sahoop@igcar.gov.in; Ananthanarayanan, R.; Murali, N. [Real Time Systems Division, Electronics, Instrumentation and Radiological Safety Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2015-02-15

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

  3. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  4. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  5. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    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.)

  6. Measurement of the heavy water level in the fuel channels of the RA reactor - Annex 11

    International Nuclear Information System (INIS)

    Nikolic, M.

    1964-01-01

    The objective of measuring the heavy water level in the reactor channels was to verify experimentally the possibilities of reactor cooling with parallel operation of heavy water pumps od 1500 rotations/min at nominal power of 6.5 MW. Measurements were done in 2 periphery and 2 central fuel channels with pumps speed 1500, 1800 and 3000 rotations/min by a contact probe with electric resistance measuring device. precision of the measurement was ±1 cm

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  9. Analysis of an ultrasonic level device for in-core Pressurized Water Reactor coolant detection

    International Nuclear Information System (INIS)

    Johnson, K.R.

    1981-01-01

    A rigorous semi-empirical approach was undertaken to model the response of an ultrasonic level device (ULD) for application to in-core coolant detection in Pressurized Water Reactors (PWRs). An equation is derived for the torsional wave velocity v/sub t phi/ in the ULD. Existing data reduction techniques were analyzed and compared to results from use of the derived equation. Both methods yield liquid level measurements with errors of approx. 5%. A sensitivity study on probe performance at reactor conditions predicts reduced level responsivity from data at lower temperatures

  10. State-of-the-art for liquid-level measurements applied to in-vessel coolant level for nuclear reactors

    International Nuclear Information System (INIS)

    Anderson, R.L.

    1980-01-01

    The TMI-2 accident indicated that a direct indication of the liquid level in the reactor vessel would have told the operators that the core was being uncovered. This state-of-the-cost survey covered the following methods: heated thermocouple, differential pressure, ultrasonic, capacitance, microwave, time-domain reflectometry, and externally mounted radiation detectors

  11. Development of a pressurizer level compensator for use on N Reactor

    International Nuclear Information System (INIS)

    Bussell, J.H.

    1985-07-01

    The instrument described in this report has been developed to compensate the measured water level in the N Reactor pressurizer for temperature effects. N Reactor is a pressurized water nuclear reactor (PWR). The instrument is defined as a pressurizer level compensator (PLC). A pressurizer is used in a PWR to control the primary coolant pressure and provide a surge volume for primary coolant expansion and contraction. A means of compensating for water and steam density is required because of the wide range of pressure and temperature that result from different steady state and transient reactor power levels. The uncompensated level is determined by measurement of differential pressure between the top of the level measurement zone and the bottom of the level measurement zone. Temperature of the water in the pressurizer is the parameter that is used to determine the proper level compensation since water and steam density are primarily functions of temperature in this case. The PLC uses a microprocessor to calculate the compensated level from temperature and differential pressure measurements. This report includes a description of the design, development, and implementation of software and hardware that are in the PLC. 9 refs., 51 figs., 17 tabs

  12. Conceptual design of a fusion-fission hybrid reactor for transmutation of high level nuclear waste

    International Nuclear Information System (INIS)

    Qiu, L.J.; Wu, Y.C.; Yang, Y.W.; Wu, Y.; Luan, G.S.; Xu, Q.; Guo, Z.J.; Xiao, B.J.

    1994-01-01

    To assess the feasibility of the transmutation of long-lived radioactive waste using fusion-fission hybrid reactors, we are studying all the possible types of blanket, including a comparison of the thermal and fast neutron spectrum blankets. Conceptual designs of a small tokamak hybrid blanket with small inventory of actinides and fission products are presented. The small inventory of wastes makes the system safer. The small hybrid reactor system based on a fusion core with experimental parameters to be realized in the near future can effectively transmute actinides and fission products at a neutron wall loading of 1MWm -2 . An innovative energy system is also presented, including a fusion driver, fuel breeder, high level waste transmuter, fission reactor and so on. An optimal combination of all types of reactor is proposed in the system. ((orig.))

  13. Application of Emergency Action Levels from Potential Release at Research Reactor HANARO

    International Nuclear Information System (INIS)

    Kim, Jongsoo; Lee, Goan Yub; Lee, Hae Choi; Kim, Bong Suk

    2014-01-01

    Execution of the protective action promptly is possible that Emergency Action Levels (EALs) must be established for a radiological release from nuclear facility. The EALs for electric power reactor are already developed and applied to recognize an emergency situation rapidly. Recently the IAEA published the safety report including the EALs for research reactor. This paper describes the EALs to apply for a potential release pathway at the research reactor HANARO. The results of table 1 and 2 will be higher than actual because the weather condition in real situation is difference. However, the EALs applying the potential stack release, ground release and site can be useful for research reactor HANARO making the emergency declaration. The EALs at the site boundary of the table 3 can be applied to protect the off-site public

  14. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  15. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  16. Measurement and evaluation of the external radiation level at reactor Kartini

    International Nuclear Information System (INIS)

    Atok Suhartanto; Suparno

    2013-01-01

    Measurement and evaluation of external radiation level at reactor Kartini in 2012 has been done. The purpose of this activity is to know the external radiation level as a result of the radioactive or radiation source usage, toward the operational of limit condition. The measurement is using survey meter Inspector 11086, factor of calibration 0.991 mR/h, at 9 locations is: Control room area, Thermal column facilities, Demineralizer, Beamport radiography facilities, bulk shielding Deck, Subcritical facilities, Reactor hall, Deck reactor and on the surface of reactor water tank . The highest room average measurement result in 9 working areas for 12 months continuously are at the reactor tank location is between 13.05±1.09 (xlO -2 mSv/hour) to 16.80±1.40 (x10 -2 mSv/hour), and the lowest measurement result in 1 location (control room) is 0.02±0.005 (x10 -2 mSv/hour) to 0.035±0.009 (x10 -2 mSv/hour). The Kartini reactor is involved in the control area which has potentially contaminated and has radiation exposure at the level of 6 mSv/year. Radiation Protection Officer that work in interval will received radiation exposure dosage of 8.4 mSv/year. This dosage is still below the Below Dosage Value which is recommended by, BAPETEN decree No, 4, 2013 about Protection and Radiation Safety in Nuclear Energy Application at 20 mSv/year. The result of the evaluation above shows that the external radiation which occurred in each area is still below the operational of limit condition that is written on the Kartini reactor safety analysis report, on document number: C7/05/B2/LAK/2010, revision 7. So that the workplace is safe for work monitored. (author)

  17. Radiation levels in the poll surface of IEA-R1 reactor

    International Nuclear Information System (INIS)

    Pasqualetto, H.

    1978-01-01

    A theoretical model for the calculation of the radioactivity level in the pool surface of the research reactor IEA-RI (INSTITUTO DE ENERGIA ATOMICA, BRAZIL) is developed. The radioactivity is caused by radionuclides (Mainly 24 Na and 27 Mg) produced by nuclear reactions of neutrons with: a) oxygen of the water b); gaseous elements dissolved in water (Ar,N); c) structural materials of the fuel can. Considerations about expected radiation level after eventual increase of reactor power from 2 MW to 10 MW are also presented [pt

  18. Safe immobilization of high-level nuclear reactor wastes

    International Nuclear Information System (INIS)

    Ringwood, A.; Kesson, S.; Ware, N.; Hibberson, W.; Major, A.

    1979-01-01

    The advantages and disadvantages of methods of immobilizing high-level radioactive wastes are discussed. Problems include the devitrification of glasses and the occurrence of radiation damage. An alternative method of radioctive waste immobilization is described in which the waste is incorporated in the constituent minerals of a synthetic rock, Synroc. Synroc is immune from devitrification and is composed of phases which possess crystal structures identical to those of minerals which are known to have retained radioactive elements in geological environments at elevated pressures and tempertures for long periods. The composition and mineralogy of Synroc is given and the process of immobilizing wastes in Synroc is described. Accelerated leaching tests at elevated pressures and temperatures are also described

  19. Safeguards challenges of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    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

  20. PWR type reactor equipped with a primary circuit loop water level gauge

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro.

    1990-01-01

    The time of lowering a water level to less than the position of high temperature side pipeway nozzle has been rather delayed because of the swelling of mixed water level due to heat generation of the reactor core. Further, there has been a certain restriction for the installation, maintenance and adjustment of a water level gauge since it is at a position under high radiation exposure. Then, a differential pressure type water level gauge with temperature compensation is disposed at a portion below a water level gauge of a pressurizer and between the steam generator exit plenum and the lower end of the loop seal. Further, a similar water level system is disposed to all of the loops of the primary circulation circuits. In a case that the amount of water contained in a reactor container should decreased upon occurrence of loss of coolant accidents caused by small rupture and stoppage of primary circuit pumps, lowering of the water level preceding to the lowering of the water level in the reactor core is detected to ensure the amount of water. Since they are disposed to all of the loops and ensure the excess margin, reliability for the detection of the amount of contained water can be improved by averaging time for the data of the water level and averaging the entire systems, even when there are vibrations in the fluid or pressure in the primary circuit. (N.H.)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  3. Steam drum level control studies of a natural circulation multi loop reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

    2013-12-15

    The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

  4. Radionuclide compositions of spent fuel and high level waste from commercial nuclear reactors

    International Nuclear Information System (INIS)

    Goodill, D.R.; Tymons, B.J.

    1984-10-01

    This report provides information on radionuclide compositions of spent fuel and high level waste produced during reprocessing. The reactor types considered are Magnox, AGR, PWR and CFR. The activities of the radionuclides are calculated using the FISPIN code. The results are presented in a form suitable for radioactive waste management calculations. (author)

  5. Bioaugmentation of Syntrophic Acetate-Oxidizing Culture in Biogas Reactors Exposed to Increasing Levels of Ammonia

    Science.gov (United States)

    Westerholm, Maria; Levén, Lotta

    2012-01-01

    The importance of syntrophic acetate oxidation for process stability in methanogenic systems operating at high ammonia concentrations has previously been emphasized. In this study we investigated bioaugmentation of syntrophic acetate-oxidizing (SAO) cultures as a possible method for decreasing the adaptation period of biogas reactors operating at gradually increased ammonia concentrations (1.5 to 11 g NH4+-N/liter). Whole stillage and cattle manure were codigested semicontinuously for about 460 days in four mesophilic anaerobic laboratory-scale reactors, and a fixed volume of SAO culture was added daily to two of the reactors. Reactor performance was evaluated in terms of biogas productivity, methane content, pH, alkalinity, and volatile fatty acid (VFA) content. The decomposition pathway of acetate was analyzed by isotopic tracer experiments, and population dynamics were monitored by quantitative PCR analyses. A shift in dominance from aceticlastic methanogenesis to SAO occurred simultaneously in all reactors, indicating no influence by bioaugmentation on the prevailing pathway. Higher abundances of Clostridium ultunense and Tepidanaerobacter acetatoxydans were associated with bioaugmentation, but no influence on Syntrophaceticus schinkii or the methanogenic population was distinguished. Overloading or accumulation of VFA did not cause notable dynamic effects on the population. Instead, the ammonia concentration had a substantial impact on the abundance level of the microorganisms surveyed. The addition of SAO culture did not affect process performance or stability against ammonia inhibition, and all four reactors deteriorated at high ammonia concentrations. Consequently, these findings further demonstrate the strong influence of ammonia on the methane-producing consortia and on the representative methanization pathway in mesophilic biogas reactors. PMID:22923397

  6. The Influence of RSG-GAS Primary Pump Operation Concerning the Rise Water Level of Reactor Pool in 15 MW Reactor Power

    International Nuclear Information System (INIS)

    Djunaidi

    2004-01-01

    The expansion of air volume in the delay chamber shows in rise water level of reactor pool during the operation. The rises of water level in the reactor pool is not quite from the expansion of air volume in the delay chamber, but some influence the primary pump operation. The purpose evaluated of influence primary pump is to know the influence primary pump power concerning the rise water level during the reactor operation. From the data collection during 15 MW power operation in the last core 42 the influence of primary pump operation concerning the rise water level in the reactor pool is 34.48 % from the total increased after operation during 12 days. (author)

  7. Analysis of Radioactivity Contamination Level of Kartini Reactor Efluen Gas to the Environment

    International Nuclear Information System (INIS)

    Suratman; Purwanto; Aminjoyo, S

    1996-01-01

    The analysis of radioactivity contamination level of Kartini reactor efluen gas to the environment has been done from 13-10-'95 until 8-2-'96. The aim of this research is to determine the radioactivity contamination level on the environment resulted from the release of Kartini reactor efluen gas and other facilities at Yogyakarta Nuclear Research Centre through stack. The analysis methods is the student t-test, the first count factor test and the gamma spectrometry. The gas sampling were carried out in the stack reactor, reactor room, environment and in other room for comparison. Efluen gas was sucked through a filter by a high volume vacuum pump. The filter was counted for beta, gamma and alpha activities. The radioactivity contamination level of the efluen gas passing through the stack to the environment was measured between 0.57 - 1.34 Bq/m3, which was equal to the airborne radioactivity in environment between 0.69 - 1.12 Bq/m3. This radioactivity comes from radon daughter, decay products result from the natural uranium and thorium series of the materials of the building

  8. Evaluation of strategies for end storage of high-level reactor fuel; Vurdering av strategier for sluttlagring av hoeyaktivt reaktorbrensel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized.

  9. Process and device for measuring the level in a reactor pressure vessel of a boiling water reactor

    International Nuclear Information System (INIS)

    Walleser, A.

    1987-01-01

    The differential pressure is measured between the lower space filled with liquid and a comparison column which is connected to the upper part space filled with steam. From this measurement and the liquid and steam densities in the pressure vessel and in the comparison column and the effects of flow at the pressure sampling positions, the level is determined in an evaluation unit. To determine the densities of liquids and steam, the reactor pressure or the change of pressure with time for transient processes is measured. The density of the comparison column is determined by temperature measurement. The effects of flow are determined by flow measurements. All the measurements are taken to an evaluation unit. (orig./HP) [de

  10. Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

    International Nuclear Information System (INIS)

    Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

    2005-01-01

    The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

  11. Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

    International Nuclear Information System (INIS)

    Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

    2014-01-01

    Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

  12. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

    2008-10-15

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

  13. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    International Nuclear Information System (INIS)

    Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

    2008-01-01

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  15. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1992-01-01

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

  16. A review of fast reactor program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

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

  17. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    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

  18. Nuclear reactor technology progress report, vol. 4

    International Nuclear Information System (INIS)

    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.)

  19. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  20. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    International Nuclear Information System (INIS)

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  1. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

    International Nuclear Information System (INIS)

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

    This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

  2. Return momentum effect on reactor coolant water level distribution during mid-loop conditions

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Yang, Jae Young; Park, Goon Cherl

    2001-01-01

    An accurate prediction of the Reactor Coolant System( RCS) water level is of importance in the determination of the allowable operating range to ensure safety during mid-loop operations. However, complex hydrualic phenomena induced by the Shutdown Cooling System (SCS) return momentum causes different water levels from those in the loop where the water level indicators are located. This was apparently observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level distribution, a model using a one-dimensional momentum and energy conservation for cylindrical channel, hydraulic jump in operating cold leg, water level build-up at the Reactor Vessel (RV) inlet nozzle, Bernoulli constant in downcomer region, and total water volume conservation has been developed. The model predicts the RCS water levels at various RCS locations during the mid-loop conditions and the calculation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs, in conjuction with the pressure drop throughout the RCS, is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels

  3. Measurement of nuclear reactor noise at low power levels; Merenje nuklearnog reaktorskog suma na malim snagama

    Energy Technology Data Exchange (ETDEWEB)

    Velickovic, Lj; Petrovic, M [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1968-11-15

    Spatial and time dependence relation of neutrons from fission and other reactions in the reactor core result in non Poisson fluctuations of neutron density. Analytical formula developed for auto-correlation function includes physical parameters which characterize time behaviour of neutrons in the reactor system. Since auto-correlation function can be easily measured, it is a useful tool for experimental determination of these parameters. Noise from the ionization chamber was measured and analyzed by a digital computer. measurements were analyzed completely in the time domain (auto-correlation functions). This enabled separating the noise caused by neutron detection from the noise from neutron density fluctuation in the reactor. All the results can be analyzed by spatial independent reactor theory. Physical analysis of reactor noise was limited to determination of {beta}/l ratio from auto-correlation measurements at 0.5 W power level (RB reactor in Vinca). Three different reactor core lattices were analyzed (lattice pitch 8 cm, 11.3 cm and 14 cm). It was shown that parameter {beta}/l could be determined from auto-correlation measurements of neutron density with high precision (few percents) Prostorna i vremenska povezanost neutrona koji nastaju u fisiji i drugim procesima koji se desavaju u reaktoru dovodi do ne Poisson-ovih fluktuacija neutronske gustine. Analiticka formula, razvijena za autokorelacionu funkciju ovih fluktuacija, sadrzi fizicke parametre koji karakterisu vremensko ponasanje neutrona u reaktorskom sistemu. Kako autokorelaciona funkcija moze lako da se meri, ona je korisno sredstvo za eksperimentalno odredjivanje ovih parametara. Sum iz jonizacione komore digitalno je meren i analiziran u digitalnom racunaru. Merenja su kompletno analizirana u vremenskom domenu (autokorelacione funkeije). To je olaksalo razdvajanje suma izazvanog procesom neutronske detekcije od suma koji potice od fluktuacija neutronske gustine u reaktorskom sistemu. Svi rezultati

  4. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

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

    International Nuclear Information System (INIS)

    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

  6. System and method for determining coolant level and flow velocity in a nuclear reactor

    Science.gov (United States)

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  7. Frequency response testing at Experimental Breeder Reactor II using discrete-level periodic signals

    International Nuclear Information System (INIS)

    Rhodes, W.D.; Larson, H.A.

    1990-01-01

    The Experimental Breeder Reactor 2 (EBR-2) reactivity-to-power frequency-response function was measured with pseudo-random, discrete-level, periodic signals. The reactor power deviation was small with insignificant perturbation of normal operation and in-place irradiation experiments. Comparison of results with measured rod oscillator data and with theoretical predictions show good agreement. Moreover, measures of input signal quality (autocorrelation function and energy spectra) confirm the ability to enable this type of frequency response determination at EBR-2. Measurements were made with the pseudo-random binary sequence, quadratic residue binary sequence, pseudo-random ternary sequence, and the multifrequency binary sequence. 10 refs., 7 figs., 3 tabs

  8. Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

    International Nuclear Information System (INIS)

    Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

    2010-01-01

    Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

  9. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    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. Enhanced xylene removal by photocatalytic oxidation using fiber-illuminated honeycomb reactor at ppb level

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Yi-Ting [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Yu, Yi-Hui [Department of Civil Engineering, National Taiwan University, Taipei 106, Taiwan (China); Nguyen, Van-Huy [Department of Chemical Engineering, National Taiwan University of Science and Technology, Taipei 10617, Taiwan (China); Lu, Kung-Te [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Wu, Jeffrey Chi-Sheng, E-mail: cswu@ntu.edu.tw [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Chang, Luh-Maan [Department of Civil Engineering, National Taiwan University, Taipei 106, Taiwan (China); Kuo, Chi-Wen [Taiwan Semiconductor Manufacturing Company, Hsinchu 30078, Taiwan (China)

    2013-11-15

    Graphical abstract: We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. -- Highlights: • The combination of optical fiber and honeycomb significantly enhanced the performance of VOCs photodegradation. • The removal efficiency of m-xylene is enhanced to 96.5% as compared to 22.0% for UV irradiation alone. • Fiber-illuminated honeycomb reactor is the first step toward an industrial-scale technology on the removal of xylene. -- Abstract: The removal of volatile organic compounds (VOCs) at ppb level is one of the most critical challenges in clean rooms for the semiconductor industry. Photocatalytic oxidation is an innovative and promising technology for ppb-level VOCs degradation. We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. Using the FIHR with Mn-TiO{sub 2} photocatalyst not only increased the m-xylene removal efficiency, but also increased the CO{sub 2} selectivity. Interestingly, Mn-TiO{sub 2} in FIHR also showed a very good reusability, 93% removal efficiency was still achieved in 72-h in reaction. Thus, the FIHR gave very high removal efficiency for xylene at ppb level under room temperature. The FIHR has great potential application in the clean room for the air purification system in the future.

  11. Enhanced xylene removal by photocatalytic oxidation using fiber-illuminated honeycomb reactor at ppb level

    International Nuclear Information System (INIS)

    Wu, Yi-Ting; Yu, Yi-Hui; Nguyen, Van-Huy; Lu, Kung-Te; Wu, Jeffrey Chi-Sheng; Chang, Luh-Maan; Kuo, Chi-Wen

    2013-01-01

    Graphical abstract: We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. -- Highlights: • The combination of optical fiber and honeycomb significantly enhanced the performance of VOCs photodegradation. • The removal efficiency of m-xylene is enhanced to 96.5% as compared to 22.0% for UV irradiation alone. • Fiber-illuminated honeycomb reactor is the first step toward an industrial-scale technology on the removal of xylene. -- Abstract: The removal of volatile organic compounds (VOCs) at ppb level is one of the most critical challenges in clean rooms for the semiconductor industry. Photocatalytic oxidation is an innovative and promising technology for ppb-level VOCs degradation. We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. Using the FIHR with Mn-TiO 2 photocatalyst not only increased the m-xylene removal efficiency, but also increased the CO 2 selectivity. Interestingly, Mn-TiO 2 in FIHR also showed a very good reusability, 93% removal efficiency was still achieved in 72-h in reaction. Thus, the FIHR gave very high removal efficiency for xylene at ppb level under room temperature. The FIHR has great potential application in the clean room for the air purification system in the future

  12. Feedback from dismantling operations (level 2) on EDF's first generation reactors

    International Nuclear Information System (INIS)

    West, J P.; Dionisio-Gomes, A.; Kus, J P.; Mervaux, P.; Bernet, P.; Dalmas, R.

    2003-01-01

    EDF's policy as regards the dismantling of the reactors that have ceased commercial operation, namely the eight power plants of the first generation and the Creys-Malville power plant, is explained. Generally speaking, prior to the year 2001, EDF had opted for the de-construction of these power plants to comply with a 'long wait' scenario, which consisted of waiting for a period of 5 to 10 years to achieve IAEA level 2 (partial release of the site), then postponing the total de-construction of the facilities for 25 to 50 years. Today, EDF has decided to undertake the total de-construction of these reactors, which have ceased commercial operation, over a period of 25 years. The purpose of this document is to present: - The reactors concerned, their background and their 'regulatory' situation, - The main operations performed and/or currently in progress, - The main elements of feedback from such operations, shedding light on the approach adopted in 2001. The installations concerned by the de-construction programme are as follows: - The 8 power plants of the first generation, which were built during the fifties and sixties and ceased commercial operation between 1973 and 1994, namely: Brennilis (industrial prototype using heavy water technology, jointly operated by EDF and CEA), the 6 power units of the NUGG type (natural uranium gas graphite) at Chinon, Saint-Laurent des Eaux and Bugey and the PWR reactor at Chooz A, - The storage silos at Saint-Laurent, where the sleeves for the fuel assemblies of reactors SLA1 and SLA2 are stored, corresponding to approximately 2000 tonnes of graphite, - The Creys-Malville reactor, FBR (fast breeder reactor) shut down in accordance with a government decision, which is currently undergoing decommissioning. At the current stage, our feedback from the dismantling operations carried out on nuclear facilities is based on (i) the work carried out or in progress that will make it possible to achieve the equivalent of IAEA level 2 in the

  13. Development of an improved-contact liquid-level probe for pressurized reactor vessels

    International Nuclear Information System (INIS)

    Kelsey, P.V. Jr.

    1982-09-01

    Electrical-conductivity-based probes for liquid level sensing show promise for pressurized water reactor environments, but have exhibited frequent bond failures at the ceramic/metal interfaces. A program to characterize and improve the interface behavior has been completed successfully, and has provided data for optimizing fabrication parameters, as well as general information on glass-to-metal bonding in a superalloy/silicate-glass system. The materials studied were Inconel X-750 and a barium silicate glass containing minor amounts of TiO 2 , CeO 2 , As 2 O 3 , Bi 2 O 3 , and Al 2 O 3

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

    International Nuclear Information System (INIS)

    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)

  15. Selection of important initiating events for Level 1 probabilistic safety assessment study at Puspati TRIGA Reactor

    International Nuclear Information System (INIS)

    Maskin, M.; Charlie, F.; Hassan, A.; Prak Tom, P.; Ramli, Z.; Mohamed, F.

    2016-01-01

    Highlights: • Identifying possible important initiating events (IEs) for Level 1 probabilistic safety assessment performed on research nuclear reactor. • Methods in screening and grouping IEs are addressed. • Focusing only on internal IEs due to random failures of components. - Abstract: This paper attempts to present the results in identifying possible important initiating events (IEs) as comprehensive as possible to be applied in the development of Level-1 probabilistic safety assessment (PSA) study. This involves the approaches in listing and the methods in screening and grouping IEs, by focusing only on the internal IEs due to random failures of components and human errors with full power operational conditions and reactor core as the radioactivity source. Five approaches were applied in listing the IEs and each step of the methodology was described and commented. The criteria in screening and grouping the IEs were also presented. The results provided the information on how the Malaysian PSA team applied the approaches in selecting the most probable IEs as complete as possible in order to ensure the set of IEs was identified systematically and as representative as possible, hence providing confidence to the completeness of the PSA study. This study is perhaps one of the first to address classic comprehensive steps in identifying important IEs to be used in a Level-1 PSA study.

  16. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

    2014-01-01

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia

  17. Neural Network with Local Memory for Nuclear Reactor Power Level Control

    International Nuclear Information System (INIS)

    Uluyol, Oender; Ragheb, Magdi; Tsoukalas, Lefteri

    2001-01-01

    A methodology is introduced for a neural network with local memory called a multilayered local output gamma feedback (LOGF) neural network within the paradigm of locally-recurrent globally-feedforward neural networks. It appears to be well-suited for the identification, prediction, and control tasks in highly dynamic systems; it allows for the presentation of different timescales through incorporation of a gamma memory. A learning algorithm based on the backpropagation-through-time approach is derived. The spatial and temporal weights of the network are iteratively optimized for a given problem using the derived learning algorithm. As a demonstration of the methodology, it is applied to the task of power level control of a nuclear reactor at different fuel cycle conditions. The results demonstrate that the LOGF neural network controller outperforms the classical as well as the state feedback-assisted classical controllers for reactor power level control by showing a better tracking of the demand power, improving the fuel and exit temperature responses, and by performing robustly in different fuel cycle and power level conditions

  18. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Mazleha Maskin; Phongsakorn, P.T.; Tonny, A.L.; Fedrick, C.M.B.; Faizal Mohamed; Mohamad Fauzi Saad; Ahmad Razali Ismail; Mohamad Puad Haji Abu

    2013-01-01

    Full-text: As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia. (author)

  19. Design considerations for economically competitive sodium cooled fast reactors

    International Nuclear Information System (INIS)

    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)

  20. The instrumentation of fast reactor

    International Nuclear Information System (INIS)

    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)

  1. Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

    International Nuclear Information System (INIS)

    Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

    1990-01-01

    A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

  2. Nuclear Data for Reactor Physics: Cross sections and level densities in the actinide region

    Directory of Open Access Journals (Sweden)

    Bernstein L.

    2010-03-01

    Full Text Available Nuclear data in the actinide region are particularly important because they are basis behind all simulations of nuclear reactor core behaviour over both long time scales (fuel depletion and waste production and short time scales (accident scenarios. Nuclear reaction cross sections must be known as precisely as possible so that core reaction rates can be accurately calculated. Although cross section measurements in this region have been widely performed, for certain nuclei, particularly those with short half lives, direct measurements are either very difficult or impossible and thus reactor simulations must rely on theoretical calculations or extrapolations from neighbouring nuclei. The greatest uncertainty in theoretical cross section calculations comes from the lack of knowledge of level densities, for which predicted values can often be incorrect by a factor of two or more. Therefore there is a strong case for a systematic experimental study of level densities in the actinide region for the purpose of a providing a stringent test of theoretical cross section calculations for nuclei where experimental cross section data are available and b for providing better estimations of cross sections for nuclei in which no cross section data are available.

  3. Effect of lattice-level adjoint-weighting on the kinetics parameters of CANDU reactors

    International Nuclear Information System (INIS)

    Nichita, Eleodor

    2009-01-01

    Space-time kinetics calculations for CANDU reactors are routinely performed using the Improved Quasistatic (IQS) method. The IQS method calculates kinetics parameters such as the effective delayed-neutron fraction and generation time using adjoint weighting. In the current implementation of IQS, the direct flux, as well as the adjoint, is calculated using a two-group cell-homogenized reactor model which is inadequate for capturing the effect of the softer energy spectrum of the delayed neutrons. Additionally, there may also be fine spatial effects that are lost because the intra-cell adjoint shape is ignored. The purpose of this work is to compare the kinetics parameters calculated using the two-group cell-homogenized model with those calculated using lattice-level fine-group heterogeneous adjoint weighting and to assess whether the differences are large enough to justify further work on incorporating lattice-level adjoint weighting into the IQS method. A second goal is to evaluate whether the use of a fine-group cell-homogenized lattice-level adjoint, such as is the current practice for Light Water Reactors (LWRs), is sufficient to capture the lattice effects in question. It is found that, for CANDU lattices, the generation time is almost unaffected by the type of adjoint used to calculate it, but that the effective delayed-neutron fraction is affected by the type of adjoint used. The effective delayed-neutron fraction calculated using the two-group cell-homogenized adjoint is 5.2% higher than the 'best' effective delayed-neutron fraction value obtained using the detailed lattice-level fine-group heterogeneous adjoint. The effective delayed-neutron fraction calculated using the fine-group cell-homogenized adjoint is only 1.7% higher than the 'best' effective delayed-neutron fraction value but is still not equal to it. This situation is different from that encountered in LWRs where weighting by a fine-group cell-homogenized adjoint is sufficient to calculate the

  4. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

  5. Root cause study on the high levels of reactor water silica and its measures for reduction in Taiwan BWRs

    International Nuclear Information System (INIS)

    Wen Tungjen; Shen Szechieh; Cheng Tingchin; Chu Charles Fang

    2009-01-01

    Water temperature in condenser hot well can change significantly with the seasons. It is not unusual for silica levels in reactor water to increase in the summer and to decrease during winter. Also it seems that silica levels in reactor water of domestic nuclear power plants increased slowly from 200 ppb to the high side above 500 ppb as condensate polishing resin non-regeneration throw away action has been adopted in recent seven years. Through both the on-line silica monitoring of condensate demineralizer and silica analysis of subsequent precoat test in reactor water clean up system, the result obtained from steam/liquid mass balance calculation indicated that an increase of reactor water silica was mainly caused by continuing equilibrium leakage from deep bed condensate demineralizers, where the ion exchange zone is periodically disturbed by resin backwashing-scrubbing operation. The fastest and effective way to reduce the silica inventory in reactor system is to operate by frequent backwashes and precoating of the reactor water clean up filter demineralizers to a lower effluent silica end point, perhaps as frequently as three or four days. Periodic replacement of the oldest and/or most heavily loaded resins could perform a radical cure to reduce silica content in reactor water. During summer season, increased filter demineralizer precoat frequency would get over the problem of system short-term silica equilibrium leakage. (author)

  6. Using level-I PRA for enhanced safety of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Ramsey, C.T.; Linn, M.A.

    1995-01-01

    The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

  7. What will we do with the low level waste from reactor decommissioning?

    International Nuclear Information System (INIS)

    Meehan, A. R.; Wilmott, S.; Crockett, G.; Watt, N. R.

    2008-01-01

    The decommissioning of the UK's Magnox reactor sites will produce large volumes of low level waste (LLW) arisings. The vast majority of this waste takes the form of concrete, building rubble and redundant plant containing relatively low levels of radioactivity. Magnox Electric Ltd (Magnox) is leading a strategic initiative funded by the Nuclear Decommissioning Authority (NDA) to explore opportunities for the disposal of such waste to suitably engineered facilities that might be located on or adjacent to the site of waste arising, if appropriate and subject to regulatory acceptance and stakeholder views. The strategic issues surrounding this initiative are described along with an update of progress with stakeholder consultations in relation to the proposed licensing of the first such facility at Hinkley Point A, which could be viewed as a test case for the development of similar disposal facilities at other nuclear sites in England and Wales. (authors)

  8. Plant level of automated control system at a NPP with RBMK reactor

    International Nuclear Information System (INIS)

    Vorob'ev, V.P.; Gorbunov, V.P.; Dmitriev, V.M.; Litvin, A.S.

    1987-01-01

    The functional structure of plant level automated control system (ACS) at NPP with RBMK-1000 reactors, its binding with the on-line control system of higher and lower levels, as well as engineering requirements to software and recommendations on composition of hardware components, are considered. NPP ACS is an organizational-engineering system consisting of computer facilities and binding aimed at solving management, economical, organizational and physical-engineering problems to control NPP more effectively. The system carries out data acquisition, preliminary processing, analysis, transmission and representation for users to accept solutions for NPP operation by operative and management personnel. The main aim of integrated NPP ACS is the control development and increase of NPP economical efficiency, the increase of electric and heat energy production, the optimization of the production distribution between units, the development of production and economic NPP control

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Study of the contamination level in the primary circuit of a pressurised water reactor

    International Nuclear Information System (INIS)

    Gomit, J.-M.

    1980-07-01

    The purpose of this study is to work out a model which allows to predict the contamination level in the primary circuit of a pressurised water reactor (PWR). We assume that the passage of fission products from fuel (UO 2 ) to water takes place in two stages: a) the fission products created in the fuel diffuse and go out into the gap; b) owing to failure of clad, fission products stored in the gap diffuse into the water. A migration constant will correspond to each of these stages (ν(C) for fuel and ν(J) for gap). We have designed two models: - an empiric model in which constants ν(C) and ν(J) are adjusted from experiments CYRANO and BOUFFON carried out by the CEA; - a theoretical model to describe the physical mechanisms of migration inside the fuel. This leads us to introduce trapping and resolution probabilities. This models lead to a theoretical definition of ν(C). We have undertaken a second qualification of the empiric model using code PROFIP 3. Application to the Tihange reactor enabled us to get a good estimation of activity in the primary circuit and the number of failed rods during the first cycle [fr

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Design of the Demineralized Water Make-up Line to Maintain the Normal Pool Water Level of the Reactor Pool in the Research Reactor

    International Nuclear Information System (INIS)

    Yoon, Hyun Gi; Choi, Jung Woon; Yoon, Ju Hyeon; Chi, Dae Young

    2012-01-01

    In many research reactors, hot water layer system (HWLS) is used to minimize the pool top radiation level. Reactor pool divided into the hot water layer at the upper part of pool and the cold part below the hot water layer with lower temperature during normal operation. Water mixing between these layers is minimized because the hot water layer is formed above cold water. Therefore the hot water layer suppresses floatation of cold water and reduces the pool top radiation level. Pool water is evaporated form the surface to the building hall because of high temperature of the hot water layer; consequently the pool level is continuously fallen. Therefore, make-up water is necessary to maintain the normal pool level. There are two way to supply demineralized water to the pool, continuous and intermittent methods. In this system design, the continuous water make-up method is adopted to minimize the disturbance of the reactor pool flow. Also, demineralized water make-up is connected to the suction line of the hot water layer system to raise the temperature of make-up water. In conclusion, make-up demineralized water with high temperature is continuously supplied to the hot water layer in the pool

  16. Fuzzy logic control of steam generator water level in pressurized water reactors

    International Nuclear Information System (INIS)

    Kuan, C.C.; Lin, C.; Hsu, C.C.

    1992-01-01

    In this paper a fuzzy logic controller is applied to control the steam generator water level in a pressurized water reactor. The method does not require a detailed mathematical mode of the object to be controlled. The design is based on a set of linguistic rules that were adopted from the human operator's experience. After off-line fuzzy computation, the controller is a lookup table, and thus, real-time control is achieved. Shrink-and-swell phenomena are considered in the linguistic rules, and the simulation results show that their effect is dramatically reduced. The performance of the control system can also be improved by changing the input and output scaling factors, which is convenient for on-line tuning

  17. Requalification of the LOFT reactor following a loss of coolant experiment (Level I)

    International Nuclear Information System (INIS)

    Cannon, J.W.

    1979-01-01

    During a Loss of Coolant Experiment (LOCE), the LOFT reactor experiences an acceleration of 10 G's and fuel cladding temperature changes at a rate of 1100 0 K/sec. These unparalleled conditions present a unique startup problem to the LOFT program: How can the integrity of the fuel be confirmed so as to minimize operation if damage has occurred. The Level I Requalification Program is designed to accomplish this. It is a progressive series of tests, designed to detect damage at the earliest possible time, and thus preclude or minimize operation if damage exists. First, fuel specialists examine the LOCE data for possible damaging conditions and the results of primary coolant sample analysis for signs of failed fuel. Second, the requalification program proceeds to a series of mechanical and physics tests

  18. Chemical conditions in the repository for low- and intermediate-level reactor waste

    International Nuclear Information System (INIS)

    Snellman, M.; Uotila, H.

    1984-01-01

    The chemical conditions in the proposed repositories for low- and intermediate-level reactor waste at Haestholmen (IVO) and Olkiluoto (TVO) have been discussed with respect to materials introduced into the repository, their possible long-term changes and interaction with groundwater flowing into the repository. The main possible groundwater-rock interactions have been discussed, as well as the role of micro-organisms, organic acids and colloids in the estimation of the barrier integrity. Experimental and theoretical studies have been performed on the basis of the natural groundwater compositions expected at the repository sites. Main emphasis is put on the chemical parameters which might influence the integrity of the different barriers in the repository as well as on the parameters which might effect the release and transport of radionuclides from the repository

  19. A level playing field-obtaining consistent cost estimates for advanced reactor designs

    International Nuclear Information System (INIS)

    Hudson, C.R.; Rohm, H.H.; Humphreys, J.R.

    1987-01-01

    A level playing field in sports is necessary to avoid a situation in which a team has an unfair advantage over its competition. Similarly, rules and guidelines for developing cost estimates can be established which, in effect, provide a level playing field whereby cost estimates for advanced power plant concepts can be presented on a consistent and equitable basis. As an example, consider the capital costs shown in Table 1. Both sets of cost are for the exact same power plant; Estimate 1 is expressed in constant dollars while Estimate 2 is presented in nominal or as-spent dollars. As shown, the costs in Table 1 are not directly comparable. Similar problems can be introduced as a result of differing assumptions in any number of parameters including the scope of the cost estimate, inflation/escalation and interest rates, contingency costs, and site location. Of course, the motivation for having consistent cost estimates is to permit comparison among various concepts. As the U.S. Department of Energy sponsors research and development work on several advanced reactor concepts in which expected cost is a key evaluation parameter, the emphasis in this particular endeavor has been in promoting the comparability of advanced reactor cost estimates among themselves and to existing power plant types. To continue with the analogy, the idea is to lay out the playing field and the rules of the contest such that each team participates in the match on an equal basis with the final score being solely determined by the inherent strengths and abilities of the teams. A description of the playing field and some of the more important rules will now be provided

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

    International Nuclear Information System (INIS)

    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)

  1. Chemical surveillance of commercial fast breeder reactors

    International Nuclear Information System (INIS)

    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

  2. Enhanced xylene removal by photocatalytic oxidation using fiber-illuminated honeycomb reactor at ppb level.

    Science.gov (United States)

    Wu, Yi-Ting; Yu, Yi-Hui; Nguyen, Van-Huy; Lu, Kung-Te; Wu, Jeffrey Chi-Sheng; Chang, Luh-Maan; Kuo, Chi-Wen

    2013-11-15

    The removal of volatile organic compounds (VOCs) at ppb level is one of the most critical challenges in clean rooms for the semiconductor industry. Photocatalytic oxidation is an innovative and promising technology for ppb-level VOCs degradation. We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. Using the FIHR with Mn-TiO2 photocatalyst not only increased the m-xylene removal efficiency, but also increased the CO2 selectivity. Interestingly, Mn-TiO2 in FIHR also showed a very good reusability, 93% removal efficiency was still achieved in 72-h in reaction. Thus, the FIHR gave very high removal efficiency for xylene at ppb level under room temperature. The FIHR has great potential application in the clean room for the air purification system in the future. Copyright © 2013 Elsevier B.V. All rights reserved.

  3. Training at the masters degree level in physics and technology of nuclear reactors in the uk

    International Nuclear Information System (INIS)

    Weaver, D.R.

    2000-01-01

    This paper discusses the current situation of university-based training for the nuclear power industry in the UK, drawing on information gathered as part of the survey for a review currently being undertaken by the Committee for Technical and Economic Studies on Nuclear Energy Development and Fuel Cycle (NDC) of the Nuclear Energy Agency (NEA) of the OECD. A particular focus will be the Physics and Technology of Nuclear Reactors MSc course at the University of Birmingham. In the past there were other similar MSc courses in the UK, but through the evolution of time the Birmingham course is now unique in its role of providing masters level training so specifically aimed at the commercial nuclear programme. Mention will, however, be made of other training at the postgraduate level elsewhere in the UK. A description is given of the need to consider a new form of relationship between industry and university in order to provide optimise the provision of masters level training. (author)

  4. Design of an optimal automatic regulator for regulating the power levels of a nuclear reactor in a wind power range

    International Nuclear Information System (INIS)

    Noori Khajavi, M.; Menhaj, M.B.; Ghofrani, M.B.

    2000-01-01

    Nuclear power reactors are, in nature nonlinear and time varying. These characteristics must be considered, if large power variations occur in their working regime. In this paper a robust optimal self-tuning regulator for regulating the power of a nuclear reactor has been designed and simulated. The proposed controller is capable of regulating power levels in a wide power range (10% to 100% power levels). The controller achieves a fast and good transient response. The simulation results show that the proposed controller outperforms the fixed optimal control recently cited in the literature for nuclear power plants

  5. Main technical options of the Jules Horowitz reactor project to achieve high flux performances and high safety level

    International Nuclear Information System (INIS)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

    2003-01-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 kW/I for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: - A safety approach based upon the defence-in-depth principle. - A strategy of generic approaches to assess experimental risks in the facility. - Internal events analysis taking into account risks linked to reactor and experiments (e.g., radioactive source-term). - Systematic consideration of external hazards (e.g., earthquake, airplane crash) and internal hazards. - Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (authors)

  6. Main technical options of the Jules Horowitz Reactor project to achieve high flux performances and high safety level

    International Nuclear Information System (INIS)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

    2003-01-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 KW/l for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: 1) A safety approach based upon the defence-in-depth principle. 2) A strategy of generic approaches to assess experimental risks in the facility. 3) Internal events analysis taking into account risks linked to reactor and experiments (eg., radioactive source-term). 4) Systematic consideration of external hazards (eg., earthquake, airplane crash) and internal hazards. 5) Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (author)

  7. A reactor-level analysis of busbar costs for US nuclear plants, 1970-2005

    International Nuclear Information System (INIS)

    Koomey, Jonathan; Hultman, Nathan E.

    2007-01-01

    We present a reactor-by-reactor analysis of historical busbar costs for 99 nuclear reactors in the United States, and compare those costs with recent projections for next-generation US reactors. We argue that cost projections far different from median historical costs require more justification than estimates that lie close to those medians. Our analysis suggests that some recent projections of capital costs, construction duration, and total operations and maintenance costs are quite low-far enough from the historical medians that additional scrutiny may be required to justify using such estimates in current policy discussions and planning

  8. Level-1 PSA to support the design of the KALIMER-600 Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Han, Sang Hoon; Kim, Tae-Woon; Jeong, Hae-Yong; Han, Seok Joong; Ahn, Kwang-Il; Yang, Joon-Eon

    2012-01-01

    A sodium-cooled fast reactor, KALIMER-600, is under development. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as a coolant. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing KALIMER-600 from the aspects of risk informed design. A preliminary level-1 internal full power PSA has been performed to evaluate the safety level and its applicability for the KALIMER-600 conceptual design. Various design alternatives are evaluated from the viewpoint of PSA in order to support the design of the KALIMER-600. Sensitivity studies are also performed to evaluate the assumptions made for the PSA. The applicability and weakness of the KALIMER-600 PSA are discussed. The technical issues to be solved in performing the PSA will be discussed. (authors)

  9. Reducing scram frequency by modifying/eliminating steam generator low-low level reactor trip setpoint for Maanshan nuclear power plant

    International Nuclear Information System (INIS)

    Yuann, R.Y.; Chiang, S.C.; Hsiue, J.K.; Chen, P.C.

    1987-01-01

    The feasibility of modification/elimination of steam generator low-low level reactor trip setpoint is evaluated by using RETRAN-02 code for the purpose of reducing scram frequency in Maanshan 3-loop pressurized water reactor. The ANS Condition II event loss of normal feedwater and condition IV event feedwater system line break are the basis for steam generator low-low level reactor trip setpoint sensitivity analysis, including various initial reactor power levels, reactivity feedback coefficients, and system functions assumptions etc., have been performed for the two basis events with steam generator low-low level reactor trip setpoint at 0% narrow range and without this trip respectively. The feasibility of modifying/eliminating current steam generator low-low level reactor trip setpoint is then determined based on whether the analysis results meet with the ANS Condition II and IV acceptance criteria or not

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  12. 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

    International Nuclear Information System (INIS)

    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

  13. Experiences from start-up and use of fail-safe reactor level meter with KNITU probes

    International Nuclear Information System (INIS)

    Badiar, S.; Sipka, J.; Vanco, P.; Slanina, Marek; Liska, L.; Gavora, D.

    2001-01-01

    Within the framework of the reconstruction of both V1 Bohunice units with WWER-440/V-230 reactors, the VUJE-SIEMENS consortium implemented measurement systems for coolant level monitoring in reactor pressure vessels with 1E qualification, resp. with the qualification of category 1 according to NUSS RG1.97. The solution uses KNITU 11 probes made by Russian POZIT company. In operating plants, the installation causes problems in relation to the existing technology and quality assurance system. In the phase of implementation, the most important tasks were to resolve component quality, to commission the system, and to check its performance. (Authors)

  14. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    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)

  15. Fuzzy logic control of water level in advanced boiling water reactor

    International Nuclear Information System (INIS)

    Lin, Chaung; Lee, Chi-Szu; Raghavan, R.; Fahrner, D.M.

    1995-01-01

    The feedwater control system in the Advanced Boiling Water Reactor (ABWR) is more challenging to design compared to other control systems in the plant, due to the possible change in level from void collapses and swells during transient events. A basic fuzzy logic controller is developed using a simplified ABWR mathematical model to demonstrate and compare the performance of this controller with a simplified conventional controller. To reduce the design effort, methods are developed to automatically tune the scaling factors and control rules. As a first step in developing the fuzzy controller, a fuzzy controller with a limited number of rules is developed to respond to normal plant transients such as setpoint changes of plant parameters and load demand changes. Various simulations for setpoint and load demand changes of plant performances were conducted to evaluate the modeled fuzzy logic design against the simplified ABWR model control system. The simulation results show that the performance of the fuzzy logic controller is comparable to that of the Proportional-Integral (PI) controller, However, the fuzzy logic controller produced shorter settling time for step setpoint changes compared to the simplified conventional controller

  16. Reactor Emergency Action Level Monitor: Volume 2, REALM user's reference guide: Final report

    International Nuclear Information System (INIS)

    Touchton, R.A.

    1988-09-01

    A User Manual for the Reactor Emergency Action Level Monitor (REALM) expert system prototype is provided in this volume. REALM has been designed to provide expert assistance in the identification of a nuclear power plant emergency situation and the determination of its severity. REALM has been developed as an expert system which can provide sensor interpretation and situation assessment in a real-time processing environment. In its state of development at project completion, these capabilities are used in an off-line (i.e., stand-alone, desktop) fashion to provide emergency preparedness assistance in the areas of emergency classification training and emergency exercise scenario generation. REALM also serves a prototype and stepping-stone for the possible connection to the plant for on-line use. In order to distinguish the off-line system (now complete) from the on-line system (now moving from a research prototype to an installed system), the term ''REALM'' is used to indicate the on-line version, with users in the control room, technical support center, and the emergency operations facility, The off-line version is referred to as ''uREALM.''

  17. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

  18. Basic approach to the disposal of low level radioactive waste generated from nuclear reactors containing comparatively high radioactivity

    International Nuclear Information System (INIS)

    Moriyama, Yoshinori

    1998-01-01

    Low level radioactive wastes (LLW) generated from nuclear reactors are classified into three categories: LLW containing comparatively high radioactivity; low level radioactive waste; very low level radioactive waste. Spent control rods, part of ion exchange resin and parts of core internals are examples of LLW containing comparatively high radioactivity. The Advisory Committee of Atomic Energy Commission published the report 'Basic Approach to the Disposal of LLW from Nuclear Reactors Containing Comparatively High Radioactivity' in October 1998. The main points of the proposed concept of disposal are as follows: dispose of underground deep enough not be disturb common land use (e.g. 50 to 100 m deep); dispose of underground where radionuclides migrate very slowly; dispose of with artificial engineered barrier which has the same function as the concrete pit; control human activities such as land use of disposal site for a few hundreds years. (author)

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

    International Nuclear Information System (INIS)

    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.)

  20. Metagenomics shows that low-energy anaerobic-aerobic treatment reactors reduce antibiotic resistance gene levels from domestic wastewater.

    Science.gov (United States)

    Christgen, Beate; Yang, Ying; Ahammad, S Z; Li, Bing; Rodriquez, D Catalina; Zhang, Tong; Graham, David W

    2015-02-17

    Effective domestic wastewater treatment is among our primary defenses against the dissemination of infectious waterborne disease. However, reducing the amount of energy used in treatment processes has become essential for the future. One low-energy treatment option is anaerobic-aerobic sequence (AAS) bioreactors, which use an anaerobic pretreatment step (e.g., anaerobic hybrid reactors) to reduce carbon levels, followed by some form of aerobic treatment. Although AAS is common in warm climates, it is not known how its compares to other treatment options relative to disease transmission, including its influence on antibiotic resistance (AR) in treated effluents. Here, we used metagenomic approaches to contrast the fate of antibiotic-resistant genes (ARG) in anaerobic, aerobic, and AAS bioreactors treating domestic wastewater. Five reactor configurations were monitored for 6 months, and treatment performance, energy use, and ARG abundance and diversity were compared in influents and effluents. AAS and aerobic reactors were superior to anaerobic units in reducing ARG-like sequence abundances, with effluent ARG levels of 29, 34, and 74 ppm (198 ppm influent), respectively. AAS and aerobic systems especially reduced aminoglycoside, tetracycline, and β-lactam ARG levels relative to anaerobic units, although 63 persistent ARG subtypes were detected in effluents from all systems (of 234 assessed). Sulfonamide and chloramphenicol ARG levels were largely unaffected by treatment, whereas a broad shift from target-specific ARGs to ARGs associated with multi-drug resistance was seen across influents and effluents. AAS reactors show promise for future applications because they can reduce more ARGs for less energy (32% less energy here), but all three treatment options have limitations and need further study.

  1. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Peng, C.M.; Maly, J.

    1988-01-01

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  2. Philosophy of safety evaluation on fast breeder reactor

    International Nuclear Information System (INIS)

    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.)

  3. Control of activation levels to simplify waste management of fusion reactor ferritic steel components

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Santoro, R.T.

    1983-01-01

    Activation characteristics of a material for service in the neutron flux of a fusion reactor first wall fall into three areas: waste management, reactor maintenance and repair, and safety. Of these, the waste management area is the most likely to impact the public acceptance of fusion reactors for power generation. The decay of the activity in steels within tens of years could lead to simplified waste disposal or possibly even to materials recycle. Whether or not these can be achieved will be controlled by (1) selection of alloying elements, (2) control of critical impurity elements, and (3) control of cross contamination from other reactor components. Several criteria can be used to judge the acceptability of potential alloying elements in iron, and to define the limits on content of critical impurity elements. One approach is to select and limit alloying additions on the basis of the activity. If material recycle is a goal, N, Al, Ni, Cu, Nb, and Mo must be excluded. If simplified waste storage by shallow land burial is the goal, regulations limit the concentration of only a few isotopes. For first-wall material that will be exposed to 9 MW-y/m 2 service, allowable initial concentration limits include (in at. ppM) Ni < 20,000; Mo < 3650; N < 3650, Cu < 2400; and Nb < 1.0. The other constituent elements of ferritic steels will not be limited. Possible substitutes for the molybdenum normally used to strengthen the steels include W, Ta, Ti, and V

  4. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  6. Water level measurement system in reactor pressure vessel of BWR and hydrogen concentration monitoring system for severe accident

    International Nuclear Information System (INIS)

    Kuroda, Hidehiko; Okazaki, Koki; Shiraishi, Fujio; Kenjyo, Hiroaki; Isoda, Koichiro

    2013-01-01

    TEPCO's Fukushima Daiichi Nuclear Power Station Accident caused severe accident to lose functions of many instrumentation systems. As a result, many important plant parameters couldn't be monitored. In order to monitor plant parameters in the case of severe accident, new instrumentation systems available in the severe conditions are being developed. Water level in reactor pressure vessel and hydrogen concentration in primary containment vessel are one of the most important parameters. Performance test results about water level measurement sensor and hydrogen sensor in severe environmental conditions are described. (author)

  7. Thermal hydraulic aspects of steam drum level control philosophy for the natural circulation based heavy water reactor

    International Nuclear Information System (INIS)

    Gupta, S.K.; Gaikwad, A.J.; Kumar, Rajesh

    2004-01-01

    From safety considerations advanced nuclear reactors rely more and more on passive systems such as natural circulation for primary heat removal. A natural circulation based water reactor is relatively larger in size so as to reduce flow losses and channel type for proper flow distribution. From the size of steam drum considerations it has to be multi loop but has a common inlet header. Normally the turbine follows the reactor. This paper addresses the thermal hydraulic aspects of the steam drum pressure and level control philosophy for a four drum, natural circulation based, channel type boiling water advanced reactor. Three philosophies may be followed for drum control viz. individual drum control, one control drum approach and an average of all the four drums. For drum pressure control, the steam flow to the turbine is be regulated. A single point pressure control is better than individual drum pressure control. This is discussed in the paper. But the control point has to be at a place down steam the point where all steam line from individual drum meet. This may lead to different pressure in all the four drums depending on the power produced in the respective loops. The difference in pressure cannot be removed even if the four drums are directly connected through pipes. Also the pressure control scheme with/without interconnection is discussed. For level, the control of individual drum may not be normally possible because of common inlet header. As the frictional pressure drops in the large diameter downcomers are small as compared to elevation pressure drops, the level in all the steam drum tend to equalize. Consequently a single representative drum level may be chosen as a control variable for controlling level in all the four drums. But in case, where all the four loops are producing different powers and single point pressure control is effective, the scheme may not work satisfactorily. the level in a drum may depend on the power produced in the loop

  8. A level playing field: Obtaining consistent cost estimates for advanced reactor designs

    International Nuclear Information System (INIS)

    Hudson, C.R. II; Rohm, H.H.; Humphreys, J.R. Jr.

    1987-01-01

    Rules and guidelines for developing cost estimates are given which provide a means for presenting cost estimates for advanced concepts on a consistent and equitable basis. For advanced reactor designs, the scope of a cost estimate includes the plant capital cost, the operating and maintenance cost, the fuel cycle cost, and the cost of decommissioning. Each element is subdivided as is necessary to provide a common reporting format for all power plant concepts. The total generation cost is taken to be a suitable choice for a summary figure of merit. To test the application of the rules and guidelines as well as developing reference costs for current technologies, several different sized coal and pressurized water reactor plant cost estimates have been prepared

  9. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  10. Shutdown radiation level and man-rem control for water cooled reactors

    International Nuclear Information System (INIS)

    Cripps, S.J.; Regan, J.D.

    1978-01-01

    The importance of controlling the formation and subsequent deposition of active corrosion products (crud) is highlighted as a method of reducing occupational exposure. A semi-empirical model is described and used to predict the effectiveness of various methods of crud control. The relative merits of reactor coolant clean-up techniques including ion-exchange and electromagnetic filtration are assessed in terms of man-rem savings and associated cost penalties. (author)

  11. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  13. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  14. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  15. Cyber security level assignment for research reactor digital instrumentation and control system architecture using concept of defense in depth

    International Nuclear Information System (INIS)

    Shin, Jin Soo; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jaek Wan

    2012-01-01

    Due to recent aging of the analog instrumentation of many nuclear power plants (NPPs) and research reactors, the system reliability decreases while maintenance and testing costs increase. In addition, it is difficult to find the substitutable analog equipment s due to obsolescence. Therefore, the instrumentation and control (I and C) systems have changed from analog system to digital system due to these facts. With the introduction of digital systems, research reactors are forced to care for the problem of cyber attacks because I and C systems have been digitalized using networks or communication systems. Especially, it is more issued at research reactors due to the accessibility of human resources. In the real world, an IBM researcher has been successful in controlling the software by penetrating a NPPs network in U.S. on July 2008 and acquiring the control right of nuclear facilities after one week. Moreover, the malignant code called 'stuxnet' impaired the nearly 1,000 centrifugal separators in Iran according to an IAEA report. The problem of cyber attacks highlights the important of cyber security, which should be emphasized. Defense.in.depth (DID) is a significant concept for the cyber security to work properly. DID institutes and maintains a hardy program for critical digital asset (CDA) by implementing multiple security boundaries. In this work, we assign cyber security levels to a typical digital I and C system using DID concept. This work is very useful in applying the concept of DID to nuclear industry with respect to cyber security

  16. Cyber security level assignment for research reactor digital instrumentation and control system architecture using concept of defense in depth

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Soo; Heo, Gyun Young [Kyung Hee University, Seoul (Korea, Republic of); Son, Han Seong [Joongbu Univ., Chungnam (Korea, Republic of); Kim, Young Ki; Park, Jaek Wan [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Due to recent aging of the analog instrumentation of many nuclear power plants (NPPs) and research reactors, the system reliability decreases while maintenance and testing costs increase. In addition, it is difficult to find the substitutable analog equipment s due to obsolescence. Therefore, the instrumentation and control (I and C) systems have changed from analog system to digital system due to these facts. With the introduction of digital systems, research reactors are forced to care for the problem of cyber attacks because I and C systems have been digitalized using networks or communication systems. Especially, it is more issued at research reactors due to the accessibility of human resources. In the real world, an IBM researcher has been successful in controlling the software by penetrating a NPPs network in U.S. on July 2008 and acquiring the control right of nuclear facilities after one week. Moreover, the malignant code called 'stuxnet' impaired the nearly 1,000 centrifugal separators in Iran according to an IAEA report. The problem of cyber attacks highlights the important of cyber security, which should be emphasized. Defense.in.depth (DID) is a significant concept for the cyber security to work properly. DID institutes and maintains a hardy program for critical digital asset (CDA) by implementing multiple security boundaries. In this work, we assign cyber security levels to a typical digital I and C system using DID concept. This work is very useful in applying the concept of DID to nuclear industry with respect to cyber security.

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

    International Nuclear Information System (INIS)

    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.)

  18. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    International Nuclear Information System (INIS)

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events

  19. The contribution of a small TRIGA university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Bastuerk, M.; Boeck, H.

    2002-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27A, installed at the pulsed neutron source ISIS at RAL are mentioned. The second topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development

  20. The contribution of a small triga university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Weber, H.W.

    2001-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27 A, installed at the pulsed neutron source ISIS at RAL will be mentioned. The second topic treated in this paper shows the international co-operation in the field of superconductors. This research work is carried out under two European TMR-Network programs. The third topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development. (orig.)

  1. A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

  2. Detailed description of an SSAC at the facility level for research reactors

    International Nuclear Information System (INIS)

    Jones, R.J.

    1984-09-01

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in a research reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  3. Control of activation levels to simplify waste management of fusion reactor ferritic steel components

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Santoro, R.T.

    1984-01-01

    The objective of this work is to examine the restrictions placed on the composition of steels to allow simplified waste management after service in a fusion reactor first wall. Decay of steel activity within tens of years could simplify waste disposal or even permit recycle. For material recycle, N, Al, Ni, Cu, Nb, and Mo must be excluded. For shallow land burial, initial concentration limits include (in at. ppM) Ni, <20,000; Mo, <3650; N, <3650; Cu, <2400; and Nb, <1.0. Other constituents of steels will not be limited

  4. Determination of the exposition rapidity in the level 49.90 of the reactor building for the decrease in the water level of the spent fuel pool

    International Nuclear Information System (INIS)

    Mijangos D, Z. E.; Herrera H, S. F.; Cruz G, M. A.; Amador C, C.

    2014-10-01

    The fuel assemblies storage in the nuclear power plant of Laguna Verde (NPP-L V) represents a crucial aspect, due to the generated dose by the decay heat of the present radio-nuclides in the assemblies retired of the reactor core, after their useful life. These spent assemblies are located inside the spent fuel pool (SFP), in the level 49.90 m in the Reload Floor of the Reactor building of NPP-L V. This leads to the protection at personnel applying the ALARA (As Low As Reasonably Achievable) criteria, fulfilling the established dose criteria by the Regulator Body the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). Considering the loss scenario of the cooling system of the SFP, in which the SFP water vaporizes, is important to know the water level in which the limit of effective dose equivalent is fulfilled for the personnel. Also, is important for the instrumentation of the SFP, for the useful life of the same instruments. In this work is obtained the exposition rapidity corresponding to different water levels of SFP in the Reload Floor of NPP-L V, to identify the minimum level of water where the limit of effective dose equivalent is fulfilled of 25 rem s to the personnel, established in the Article 48 of the General Regulation of Radiological Safety of CNSNS and the Chapter 50 Section 67 of the 10-Cfr of Nuclear Regulatory Commission in USA. The water level is also identified where the exposition rapidity is of 15 m R/hr, being the value of the set point of the area radiation monitor D21-Re-N003-1, located to 125 cm over the level 49.90 meters of the Reload Floor of NPP-L V. (Author)

  5. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

    International Nuclear Information System (INIS)

    Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  6. Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

  7. Progress and challenges in the development and qualification of multi-level multi-physics coupled methodologies for reactor analysis

    International Nuclear Information System (INIS)

    Ivanov, K.; Avramova, M.

    2007-01-01

    Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics multi-scale coupled code systems. The efforts have been focused on extending the analysis capabilities by coupling models, which simulate different phenomena or system components, as well as on refining the scale and level of detail of the coupling. This paper reviews the progress made in this area and outlines the remaining challenges. The discussion is illustrated with examples based on neutronics/thermohydraulics coupling in the reactor core modeling. In both fields recent advances and developments are towards more physics-based high-fidelity simulations, which require implementation of improved and flexible coupling methodologies. First, the progresses in coupling of different physics codes along with the advances in multi-level techniques for coupled code simulations are discussed. Second, the issues related to the consistent qualification of coupled multi-physics and multi-scale code systems for design and safety evaluation are presented. The increased importance of uncertainty and sensitivity analysis are discussed along with approaches to propagate the uncertainty quantification between the codes. The incoming OECD LWR Uncertainty Analysis in Modeling (UAM) benchmark is the first international activity to address this issue and it is described in the paper. Finally, the remaining challenges with multi-physics coupling are outlined. (authors)

  8. Progress and challenges in the development and qualification of multi-level multi-physics coupled methodologies for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, K.; Avramova, M. [Pennsylvania State Univ., University Park, PA (United States)

    2007-07-01

    Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics multi-scale coupled code systems. The efforts have been focused on extending the analysis capabilities by coupling models, which simulate different phenomena or system components, as well as on refining the scale and level of detail of the coupling. This paper reviews the progress made in this area and outlines the remaining challenges. The discussion is illustrated with examples based on neutronics/thermohydraulics coupling in the reactor core modeling. In both fields recent advances and developments are towards more physics-based high-fidelity simulations, which require implementation of improved and flexible coupling methodologies. First, the progresses in coupling of different physics codes along with the advances in multi-level techniques for coupled code simulations are discussed. Second, the issues related to the consistent qualification of coupled multi-physics and multi-scale code systems for design and safety evaluation are presented. The increased importance of uncertainty and sensitivity analysis are discussed along with approaches to propagate the uncertainty quantification between the codes. The incoming OECD LWR Uncertainty Analysis in Modeling (UAM) benchmark is the first international activity to address this issue and it is described in the paper. Finally, the remaining challenges with multi-physics coupling are outlined. (authors)

  9. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  10. Estimation of CANDU reactor zone controller level by generalized perturbation method

    International Nuclear Information System (INIS)

    Kim, Do Heon; Kim, Jong Kyung; Choi, Hang Bok; Roh, Gyu Hong; Yang, Won Sik

    1998-01-01

    The zone controller level change due to refueling operation has been studied using a generalized perturbation method. The generalized perturbation method provides sensitivity of zone power to individual refueling operation and incremental change of zone controller level. By constructing a system equation for each zone power, the zone controller level change was obtained. The details and a proposed model for future work are described

  11. Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2015-11-15

    Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

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

    International Nuclear Information System (INIS)

    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)

  13. Measurement of the heavy water level in the fuel channels of the RA reactor - Annex 11; Prilog 11- Merenje nivoa teske vode u gorivnim kanalima reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1964-12-15

    The objective of measuring the heavy water level in the reactor channels was to verify experimentally the possibilities of reactor cooling with parallel operation of heavy water pumps od 1500 rotations/min at nominal power of 6.5 MW. Measurements were done in 2 periphery and 2 central fuel channels with pumps speed 1500, 1800 and 3000 rotations/min by a contact probe with electric resistance measuring device. precision of the measurement was {+-}1 cm.

  14. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  15. Achievement of the level 1 PSA in support to the CEA 2400 MWth gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Balmain, M.; Bassi, C.; Azria, P.

    2012-01-01

    Within Generation IV International Forum, the CEA has developed since 2006 a Level 1 PSA to support the design of the 2400 MWth GFR (Gas-cooled Fast Reactor). A first period, with insights published in 2008, consisted in a model with few initiators representative of medium and high pressure situations, those used for the deterministic design of the Decay Heat Removal (DHR) dedicated loops. In a second period, an iterative work reached the probabilistic targets used for generation III reactors, with prior use of normal loops, and increase of DHR reliability in high pressure conditions. The PSA team covered all the internal initiators, and supported the design of components with instrumentation and control and electrical supplies, and the shutdown operating modes of secondary, tertiary circuits, with possible re-alignment to dedicated DHR loops. Besides, the completed PSA integrated more realistic success criteria than the preliminary model and than the deterministic approach, thanks to CATHARE2 code. In case of loss of Forced Convection, the probability of success of the Natural Convection DHR was assessed by a reliability method for passive systems. The paper underlines the PSA methodology knowledge from the EDF expertise, the improvements co-developed with CEA, and the iteration design-PSA-design. (authors)

  16. Submersible microbial desalination cell for simultaneous ammonia recovery and electricity production from anaerobic reactors containing high levels of ammonia

    DEFF Research Database (Denmark)

    Zhang, Yifeng; Angelidaki, Irini

    2015-01-01

    High ammonia concentration in anaerobic reactors can seriously inhibit the anaerobic digestion process. In this study, a submersible microbial desalination cell (SMDC) was developed as an innovative method to lower the ammonia level in a continuous stirred tank reactor (CSTR) by in situ ammonia...... recovery and electricity production. In batch experiment, the ammonia concentration in the CSTR decreased from 6 to 0.7g-N/L during 30days, resulting in an average recovery rate of 80g-N/m2/d. Meanwhile, a maximum power density of 0.71±0.5W/m2 was generated at 2.85A/m2. Both current driven NH4+ migration...... and free NH3 diffusion were identified as the mechanisms responsible for the ammonia transportation. With an increase in initial ammonia concentration and a decrease in external resistance, the SMDC performance was enhanced. In addition, the coexistence of other cations in CSTR or cathode had no negative...

  17. Submersible microbial desalination cell for simultaneous ammonia recovery and electricity production from anaerobic reactors containing high levels of ammonia.

    Science.gov (United States)

    Zhang, Yifeng; Angelidaki, Irini

    2015-02-01

    High ammonia concentration in anaerobic reactors can seriously inhibit the anaerobic digestion process. In this study, a submersible microbial desalination cell (SMDC) was developed as an innovative method to lower the ammonia level in a continuous stirred tank reactor (CSTR) by in situ ammonia recovery and electricity production. In batch experiment, the ammonia concentration in the CSTR decreased from 6 to 0.7 g-N/L during 30 days, resulting in an average recovery rate of 80 g-N/m(2)/d. Meanwhile, a maximum power density of 0.71±0.5 W/m(2) was generated at 2.85 A/m(2). Both current driven NH4(+) migration and free NH3 diffusion were identified as the mechanisms responsible for the ammonia transportation. With an increase in initial ammonia concentration and a decrease in external resistance, the SMDC performance was enhanced. In addition, the coexistence of other cations in CSTR or cathode had no negative effect on the ammonia transportation. Copyright © 2014 Elsevier Ltd. All rights reserved.

  18. Study on treating of low-level radioactive reactor wastewater by combined membrane process (UF-RO)

    International Nuclear Information System (INIS)

    Lu Yunyun; Cao Qiru; Chen Yunming; Huang Lijuan; Bai Xiaofeng; Li Bing; Feng Liang

    2013-01-01

    According to the characteristics of radionuclide exists in the low-level radioactive reactor waste water from HFETR, we use a new combined membrane process separation technology to study the efficient treating of low-lever radioactive reactor wastewater. First, the prepared the simulated wastewater contained Cs + , Sr 2+ , CO 2+ , Ni 2+ , and Fe 3+ . Then, we sequentially investigated the pressure, ion concentration, pH value and EDTA, which have effects on the desalination rate of membrane processing metal ions in wastewater. The results show that: in the condition of pH = 7, and added 0.15 mol/L EDTA, the simulated wastewater separated by UF-RO, desalination rates of Cs + , Sr 2+ , CO 2+ , Ni 2+ and Fe 3+ are all above 95%; In the subsequent trials, adding 0.15 mol/L EDTA into the radioactive residuary solution, and then treating by UF-RO-RO, the decontamination efficiency can reach 95.7%. (authors)

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

    International Nuclear Information System (INIS)

    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

  20. Core design experience of WWER-440 reactors when they working on increased power level

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.; Melenchuk, I.

    2015-01-01

    The Kola NPP continues commercial operation of 2nd generation fuel (FA-2) and trial operation of 3rd generation fuel (FA-3), which has a number of design features providing the best operational characteristics. This report gives the results of VVER-440 core operation with FA-2 and FA-3 with enrichment increased up to 4.87%, and at the power level uprated to 107% of nominal power level. Brief analysis of obtained data is carried out. Peculiarities and techniques of developing loading patterns with new types of nuclear fuel for operation at the uprated power level are reviewed. (authors)

  1. An analysis of the water-level monitoring system for a boiling-water reactor

    International Nuclear Information System (INIS)

    Carlson, R.W.; Belblidia, L.A.; Russell, J.L. Jr.

    1985-01-01

    The water-level instrumentation system is very important to the overall safety of a BWR. This system is being monitored by the Safety Parameter Display System (SPDS) that is being installed in Georgia Power Company's Plant Hatch. One of the most significant functions of the SPDS is the comparison of redundant instrument readings and formation of the best estimate of each parameter from those readings which are consistent. When comparing water-level instrument readings, it is necessary to correct the individual readings for differences between current and calibration conditions as well as for differences between calibration conditions for the multiple instruments. This paper documents the examination of the water-level instrumentation system at Plant Hatch and presents the development of the equations that were used to determine the differences between indicated and actual water levels. (author)

  2. Action level for imported food in Japan after the reactor accident at Chernobyl

    International Nuclear Information System (INIS)

    Okubo, Takashi

    1988-01-01

    The nuclear accident at Chernobyl in Apr. 1986 caused a widespread release of radionuclides to environment. As a result of food movement in international trade, it was necessary to decide action level of radionuclides for food imported in Japan. The action level was derived from the following basic principle: Dose equivalent should be less than one third of 0.5 rem/year for whole body exposure. Assuming that the composition of representative radionuclides ( 90 Sr, 134 Cs and 137 Cs) in imported food are equal to those of fallout in Japan and consumption of internal food products reduces total intake of radionuclides to 35 %, action level indicated by sum of 134 Cs and 137 Cs concentrations was estimated to be 370 Bq/kg. From Nov. 1986 to Sep. 1987, it was observed that twenty samples in imported food contained radioactivity exceeding the action level. (author)

  3. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  4. Determination of trace level anions in reactor cooling water by ion chromatography using a resin of low capacity

    International Nuclear Information System (INIS)

    1988-01-01

    In the field of nuclear technology, IC has been found to be one of the most versatile and efficient analytical technique because of its ability to provide a fast and sensitive analysis of anions. In this work, a separater column packed with a resin of very low capacity was used with a concentration column for the determination of low level of anions present as traces in reactor cooling water. The results of retention times and detection limits were determined using 2.4 mM Na2CO3 / 3mM NaHCO3 mixture as eluent. The interferences of cations and anions such as (Ca)2+, (UO2)2+, (SO4)2+, have been investigated

  5. The C language auto-generation of reactor trip logic caused by steam generator water level using CASE tools

    International Nuclear Information System (INIS)

    Kim, Jang Yeol; Lee, Jang Soo

    1999-01-01

    The purpose is to produce a model of nuclear reactor trip logic caused by the steam generator water level of Wolsung 2/3/4 unit through an activity chart and a statechart and to produce C language automatically using statechart-based formalism and statemate MAGNUM toolset suggested by David Harel Formalism. It was worth attempting auto-generation of C language through we manually made Software Requirement specification(SRS) for safety-critical software using statechart-based formalism. Most of the phase of the software life-cycle except the software requirement specification of an analysis phase were generated automatically by Computer Aided Software Engineering(CASE) tools. It was verified that automatically produced C language has high productivity, portability, and quality through the simulation. (Author). 6 refs., 6 figs

  6. Some problems of software development for the plant-level automated control system of NPPs with the RBMK reactors

    International Nuclear Information System (INIS)

    Gorbunov, V.P.; Egorov, A.K.; Isaev, N.V.; Saprykin, E.M.

    1987-01-01

    Problems on development and operation of automated control system (ACS) software of NPPs with the RBMK reactors are discussed. The ES computer with large on-line storage (not less than 1 Mbite) and fast response (not less than 300.000 of operations per a second) should enter the ACS composition. Several program complexes are used in the NPP ACS. The programs collected into the EhNERGIYa library are used to provide central control system operation. The information-retrival system called the Fuel file is used to automate NPP fuel motion account, as well as to estimate efficiency of fuel application, to carry out calculations of a fuel component of electric and heat energy production cost. The automated information system for unit operation efficiency analysis, which solves both plant and unit-level problems, including engineering and economical factors and complexing of operation parameter bank, is under trial operation

  7. Output feedback dissipation control for the power-level of modular high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Dong, Z.

    2011-01-01

    Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR) is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis. (author)

  8. Output Feedback Dissipation Control for the Power-Level of Modular High-Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2011-11-01

    Full Text Available Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis.

  9. Analysis of convergence of uncertainty and important factors affecting uncertainty in level 1 PSA for pressurized water reactors

    International Nuclear Information System (INIS)

    Shimada, Yoshio

    2002-01-01

    We analyzed how the convergence of mean core damage frequency (CDF) depends on the number of minimal cut sets, the sampling method and the random seed, using level 1 PSA models for Surry 1 and a Japanese 4 loop PWR plant. As a result, the followings were clarified: the good convergence efficiency of the latin hypercube sampling (LHS), the relationship between number of minimal cut sets and mean CDF, as well as the standard deviation and the easy method of judgment for mean CDF convergence. In addition, it was seen that the relationship between the number of probability variables (i.e. the number of basic events) and the number of samplings needed to converge for mean CDF. Analysis of important factors affecting uncertainty was also performed. As a result, it was found that the initiating events (especially loss of coolant accidents) were the dominant important factors. Finally, comparisons were made for the 95% confidence interval of the calculated results from the operating experience of the worldwide nuclear power plants with (1) the mean core damage frequency by PSA for 108 US plants and 51 Japanese plants and (2) the 95% confidence interval of the US and the Japanese Plant PSA model used in this research. As a result, it was clarified that the mean core damage frequency of almost all US pressurized and boiling light water reactors in the US was in the 90% confidence interval calculated from the operating experience of the nuclear power plants (PWRs and BWRs) in the world, but that of those reactors in Japan was smaller then that level. (author)

  10. Analysis of convergence of uncertainty and important factors affecting uncertainty in level 1 PSA for pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Yoshio [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2002-09-01

    We analyzed how the convergence of mean core damage frequency (CDF) depends on the number of minimal cut sets, the sampling method and the random seed, using level 1 PSA models for Surry 1 and a Japanese 4 loop PWR plant. As a result, the followings were clarified: the good convergence efficiency of the latin hypercube sampling (LHS), the relationship between number of minimal cut sets and mean CDF, as well as the standard deviation and the easy method of judgment for mean CDF convergence. In addition, it was seen that the relationship between the number of probability variables (i.e. the number of basic events) and the number of samplings needed to converge for mean CDF. Analysis of important factors affecting uncertainty was also performed. As a result, it was found that the initiating events (especially loss of coolant accidents) were the dominant important factors. Finally, comparisons were made for the 95% confidence interval of the calculated results from the operating experience of the worldwide nuclear power plants with (1) the mean core damage frequency by PSA for 108 US plants and 51 Japanese plants and (2) the 95% confidence interval of the US and the Japanese Plant PSA model used in this research. As a result, it was clarified that the mean core damage frequency of almost all US pressurized and boiling light water reactors in the US was in the 90% confidence interval calculated from the operating experience of the nuclear power plants (PWRs and BWRs) in the world, but that of those reactors in Japan was smaller then that level. (author)

  11. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  12. Preconstruction radioactivity levels in the vicinity of the proposed Clinch River Breeder Reactor Project

    International Nuclear Information System (INIS)

    1984-05-01

    Routine samples of ground water, river water, and bottom sediment were collected from the Clinch River in 1983 in the preconstruction-construction phase of the CRBRP environmental radiological monitoring program. The water samples analyzed for iodine-131 yielded only a slight indication of the presence of I-131 at levels below the nominal lower limit of detection of 0.5 pCi/L. The only significant radioisotopes identified in sediment samples were 137 Cs, 60 Co, and the naturally occurring 40 K. The results for 137 Cs vary from 2.2 to 10.1 pCi/g (dry weight), while the results for 60 Co range from 0.35 to 1.2 pCi/g (dry weight). With the exception of tritium, no significant radioactivity was detected in ground or surface water at the CRBRP site. Tritium concentrations ranging from 12,667 to 12,823 pCi/L were found in samples of surface water taken from the Clinch River below Melton Hill Dam while samples taken at the dam exhibited tritium levels from 28 to 942 pCi/L. These elevated tritium levels in the Clinch River below Melton Hill Dam are attributable to DOE operations at Oak Ridge. The external gamma radiation levels measured at the CRBRP site averaged 17.4 +- 3.2 mR/quarter for 1983. This is consistent with levels measured at TVA's nonoperating nuclear power plant construction sites. 3 figures, 8 tables

  13. 77 FR 20059 - License Amendment To Increase the Maximum Reactor Power Level, Florida Power & Light Company...

    Science.gov (United States)

    2012-04-03

    ... reduce and mitigate salinity levels in groundwater; operational changes to the PTN cooling canal system... approving the proposed expansion of PTN Units 3 and 4 based on compliance with conditions required by the... restrooms would be used during plant modifications. Therefore, land use conditions would not change at the...

  14. Current issues in the management of low- and intermediate-level radioactive wastes from Ontario Hydro's CANDU reactors

    International Nuclear Information System (INIS)

    Krasznai, J.P.; Vaughan, B.R.; Williamson, A.S.

    1990-01-01

    Nuclear generating stations (NGSs) in Canada are operated by utilities in Ontario, Quebec, and New Brunswick. Ontario Hydro, with a committed nuclear program of 13,600 MW(electric) is the major producer of CANDU pressurized heavy-water reactor (PHWR) low- and intermediate-level radioactive wastes. All radioactive wastes with the exception of irradiated fuel are processed and retrievably stored at a centralized facility at the Bruce Nuclear Power Development site. Solid-waste classifications and annual production levels are given. Solid-waste management practices at the site as well as the physical, chemical, and radiochemical characteristics of the wastes are well documented. The paper summarizes types, current inventory, and estimated annual production rate of liquid waste. Operation of the tritium recovery facility at Darlington NGS, which removes tritium from heavy water and produces tritium gas in the process, gives rise to secondary streams of tritiated solid and liquid wastes, which will receive special treatment and packaging. In addition to the treatment of radioactive liquid wastes, there are a number of other important issues in low- and intermediate-level radioactive waste management that Ontario Hydro will be addressing over the next few years. The most pressing of these is the reduction of radioactive wastes through in-station material control, employee awareness, and improved waste characterization and segregation programs. Since Ontario Hydro intends to store retrievable wastes for > 50 yr, it is necessary to determine the behavior of wastes under long-term storage conditions

  15. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  16. Determination of the level of water in the core of reactors PWR using neutron detectors signal ex core; Determinacion del nivel del agua del nucleo de reactores PWR usando la senal de detectores neutronicos excore

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, A.; Abarca, A.; Miro, R.; Verdu, G.

    2014-07-01

    The level of water from the core provides relevant information of the neutronic and thermal hydraulic of the reactor as the power, k EFF and cooling capacity. In fact, this level monitoring can be used for prediction of LOCA and reduction of cooling that can cause damage to the core. There are several teams that measure a variety of parameters of the reactor, as opposed to the level of the water of the core. However, the detectors 'excore' measure fast neutrons which escape from the core and there are studies that demonstrate the existence of a relationship between them and the water level of the kernel due to the water shield. Therefore, a methodology has been developed to determine this relationship, using the Monte Carlo method using the MCNP code and apply variance reduction techniques based on the attached flow that is obtained using the method of discrete ordinates using code TORT. (Author)

  17. Raising the four downcomers in the reactor aluminium tank of the FRJ-2 research reactor as an example of the execution of complicated work in the region of high radiation levels

    International Nuclear Information System (INIS)

    Nickel, M.; Schmitz, J.; Wolters, J.

    1975-02-01

    As a result of the planned power increase from 15 MW to 25 MW, a new emergency cooling system had to be installed in the research reactor FRJ-2 of the KFA Juelich, which called for an extension of the four standpipes in the reactor tank by 57 mm. Due to the high radiation level in the reactor tank, new techniques had to be found allowing aluminium rings of corresponding height to be welded onto the top part of the standpipes by remotecontrolled welding; moreover, the welded parts were then to be protected by a bandage made of high-quality steel. The development work was carried out in the KFA and this report gives an account of the technique applied and the results obtained. (author)

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

    International Nuclear Information System (INIS)

    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

  19. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1979-01-01

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

  20. Modeling and monitoring cyclic and linear volatile methylsiloxanes in a wastewater treatment plant using constant water level sequencing batch reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wang, De-Gao, E-mail: degaowang@dlmu.edu.cn; Du, Juan; Pei, Wei; Liu, Yongjun; Guo, Mingxing

    2015-04-15

    The fate of cyclic and linear volatile methylsiloxanes (VMSs) was evaluated in a wastewater treatment plant (WWTP) using constant water level sequencing batch reactors from Dalian, China. Influent, effluent, and sewage sludge samples were collected for seven consecutive days. The mean concentrations of cyclic VMSs (cVMSs) in influent and effluent samples are 1.05 μg L{sup −1} and 0.343 μg L{sup −1}; the total removal efficiency of VMSs is > 60%. Linear VMS (lVMS) concentration is under the quantification limitation in aquatic samples but is found in sludge samples with a value of 90 μg kg{sup −1}. High solid-water partition coefficients result in high VMS concentrations in sludge with the mean value of 5030 μg kg{sup −1}. No significant differences of the daily mass flows are found when comparing the concentration during the weekend and during working days. The estimated mass load of total cVMSs is 194 mg d{sup −1} 1000 inhabitants{sup −1} derived for the population. A mass balance model of the WWTP was developed and derived to simulate the fate of cVMSs. The removal by sorption on sludge increases, and the volatilization decreases with increasing hydrophobicity and decreasing volatility for cVMSs. Sensitivity analysis shows that the total suspended solid concentration in the effluent, mixed liquor suspended solid concentration, the sewage sludge flow rate, and the influent flow rate are the most influential parameters on the mass distribution of cVMSs in this WWTP. - Highlights: • A mass balance model for siloxanes was developed in sequencing batch reactor. • Total suspended solid in effluent has the most influence on removal efficiency. • Enhancement of suspended solid removal reduces the release to aquatic environment.

  1. Update of preconstruction radioactivity levels in the vicinity of the proposed Clinch River breeder reactor project

    International Nuclear Information System (INIS)

    1981-08-01

    Routine samples of ground water, river water, and bottom sediment were collected from the Clinch River in 1977 in the preconstruction-construction phase of the CRBRP environmental radiological monitoring program. The results obtained from the analysis of these samples are similar to those reported earlier. The only significant radioisotopes identified in sediment samples were 137 Cs, 60 Co, and the naturally occurring 40 K. Other than for samples collected in November 1976, the results vary from 0.1 to 13.0 pCi/g (dry weight), with concentration generally increasing with distance downstream from CRM 24.0 to CRM 14.4. The extent to which this relationship may or may not hold below CRM 14.4 is beyond the intended scope of this program. No explanation can be given at this time for the elevated levels of 137 Cs detected in November 1976. These values ranged from 28.0 to 82.6 pCi/g (dry weight), with similar values found at CRM 24.0 above Melton Hill Dam. With the exception of tritium, no significant radioactivity was detected in ground or surface water at the CRBRP site. Tritium concentrations ranging from 368 to 5882 pCi/l were found in samples of surface water taken from the Clinch River below Melton Hill Dam while samples taken above the dam exhibited tritium levels from 56 to 368 pCi/l. These elevated tritium levels in the Clinch River below Melton Hill Dam are attributable to DOE operations at Oak Ridge

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  3. Consignment of Very Low Level Waste (VLLW) from the Winfrith dragon reactor containment building

    International Nuclear Information System (INIS)

    Shuler, K.

    2008-01-01

    The United Kingdom Atomic Energy Authority (UKAEA), CH2M Hill and AMEC are implementing innovative technical approaches in the decommissioning of redundant nuclear plant. These approaches will form the basis of lessons learned and best practices to be applied to future decommissioning work across the United Kingdom. This paper highlights the approach used for categorizing waste from the Dragon Decommissioning Project as Very Low Level Waste (VLLW), a category typically used by hospitals and laboratories for small quantities of waste contaminated with radioisotopes. (authors)

  4. Immobilization of high level nuclear reactor wastes in SYNROC: a current appraisal

    International Nuclear Information System (INIS)

    Oversby, V.M.; Ringwood, A.E.

    1981-01-01

    Results are presented for leach testing at 95 0 C and 200 0 C of SYNROC containing 9% and 20% simulated high level radioactive waste, synthetic hollandite and pervoskite samples, and natural zirconolite and pervoskite samples. Single phase synthetic minerals show much higher leach rates than natural mineral samples and polyphase SYNROC samples. Natural zirconolite samples with low radiation damage have leach rates at 200 0 C based on U which are identical to those measured on SYNROC samples. Natural zirconolites with very large accumulated α dose and radiation damage have leach rates at 200 0 C which are only 5 times higher than those of low dose samples

  5. A review of fast reactor programme in Japan

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    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

  7. Air radioactivity levels following the Fukushima reactor accident measured at the Laboratoire Souterrain de Modane, France.

    Science.gov (United States)

    Loaiza, P; Brudanin, V; Piquemal, F; Reyss, J-L; Stekl, I; Warot, G; Zampaolo, M

    2012-12-01

    The radioactivity levels in the air of the radionuclides released by the Fukushima accident were measured at the Laboratoire Souterrain de Modane, in the South-East of France, during the period 25 March-18 April 2011. Air-filters from the ventilation system exposed for one or two days were measured using low-background gamma-ray spectrometry. In this paper we present the activity concentrations obtained for the radionuclides (131)I, (132)Te, (134)Cs, (137)Cs, (95)Nb, (95)Zr, (106)Ru, (140)Ba/La and (103)Ru. The activity concentration of (131)I was of the order of 100 μBq/m(3), more than 100 times higher than the activities of other fission products. The highest activities of (131)I were measured as a first peak on 30 March and a second peak on 3-4 April. The activity concentrations of (134)Cs and (137)Cs varied from 5 to 30 μBq/m(3). The highest activity concentration recorded for Cs corresponded to the same period as for (131)I, with a peak on 2-3 April. The results of the radioactivity concentration levels in grass and mushrooms exposed to the air in the Modane region were also measured. Activity concentrations of (131)I of about 100 mBq/m(2) were found in grass. Copyright © 2012 Elsevier Ltd. All rights reserved.

  8. Performance analysis of a repository for low and intermediate level reactor waste

    International Nuclear Information System (INIS)

    Vieno, T.; Nordman, H.; Vuori, S.; Peltonen, E.

    1987-01-01

    In Finland, utilities producing nuclear energy are responsible for the management of the radioactive waste, including final disposal. As regards low and intermediate level waste, the approach has been adopted to employ the power plant sites for locations of repositories. The repositories will be excavated at the depth of about 50 to 125 m in the bedrock of the two Finnish nuclear power plant sites, Loviisa and Olkiluoto. The performance analysis presented in this paper has been carried out for the Preliminary Safety Analysis Report (PSAR) of the Olkiluoto repository. A flexible model has been developed to estimate the release of radionuclides from waste packages and their subsequent transport through the engineered barriers in the repository. Gradual degradation of the engineered barriers is accounted for by altering parameters at fixed time points. Safety margins of the disposal concept have been evaluated by including disturbed evolution scenarios in the analysis. 13 references, 10 figures, 2 tables

  9. For grasping the Pu background level in the soils of environment around Kyoto University Reactor

    International Nuclear Information System (INIS)

    Ikeda, E.; Fujikawa, Y.; Fukui, M.; Saito, M.

    2001-01-01

    In the case of emergency evaluation of the contamination level of plutonium at the environment in a short time is required. R and D of plutonium analysis in a soil is performed using ICP-MS. The detection limit of the plutonium by ICP-MS is almost same as 2mBq, which is the detection goal of α-ray spectrometry by Japan Analysis Center. It became possible to carry out the quantitative analysis of fall out plutonium at the environment in a short time of about several ten seconds. For the soils used in the analysis experiment the dry and combustion processes were found to be able to skip, which is for removing the organic compounds through the pretreatment of the specimen. (Katsuta, H.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Sakamoto, Yoshihiko

    2010-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    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)

  15. Feasibility study on vitrification of low- and intermediate-level radioactive waste from pressurized water reactors

    International Nuclear Information System (INIS)

    Park, J.K.; Song, M.J.

    1998-01-01

    In order to obtain annual generation volume and composition data for low- and intermediate-level radioactive waste (LILW), characteristics and generation trends for each waste which was produced at nuclear power plants (NPPs) in Korea were investigated. Of the three different types of melters, the platinum crucible was found to be most suitable for the performance of vitrification experiments and hence, was used to help better understand the optimal waste contents in borosilicate glass waste forms with respect to waste types. After the performance of vitrification experiments, compressive strength tests showed that the final waste glass product, containing up to 40 vol% of ashy pyrolyzed/oxidized at 400--800 C, showed good mechanical stability and homogeneity in the glass matrix. Economical assessment was performed with some considerations given for equipment having already been adopted for LILW treatment in Korea for four treatment strategies with melters selected from a technical assessment. For each strategy, the capital and the operation cost were estimated, and the disposal volume was calculated with reasonably estimated volume reduction factors with regard to waste type and treatment concept

  16. The regulation on commercial reactors and the management of high-level radioactive wastes in U.S

    International Nuclear Information System (INIS)

    Shimomura, Hidetsugu

    2013-01-01

    This article shows U.S. NRC's substantial and procedural regulations regarding commercial reactors and radioactive wastes. The commercial reactor's regulations are analyzed from an ensuring safety, and the radioactive waste' management is done from a locating a disposal site. (author)

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

    International Nuclear Information System (INIS)

    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)

  18. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    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

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  20. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  1. Contamination level of four priority phthalates in North Indian wastewater treatment plants and their fate in sequencing batch reactor systems.

    Science.gov (United States)

    Gani, Khalid Muzamil; Rajpal, Ankur; Kazmi, Absar Ahmad

    2016-03-01

    The contamination level of four phthalates in untreated and treated wastewater of fifteen wastewater treatment plants (WWTPs) and their fate in a full scale sequencing batch reactor (SBR) based WWTP was evaluated in this study. The four phthalates were diethyl phthalate (DEP), dibutyl phthalate (DBP), benzylbutyl phthalate (BBP) and diethylhexyl phthalate (DEHP). All compounds were present in untreated wastewater with DEHP being present in the highest mean concentration of 28.4 ± 5.3 μg L(-1). The concentration was in the range of 7.3 μg L(-1) (BBP) to 28.4 μg L(-1) (DEHP) in untreated wastewater and 1.3 μg L(-1) (DBP) to 2.6 μg L(-1) (DEHP) in treated wastewater. The nutrient removal process and advance tertiary treatment based WWTPs showed the highest phthalate removal efficiencies of 87% and 93%, respectively. The correlation between phthalate removal and conventional performance of WWTPs was positive. Fate analysis of these phthalates in a SBR based WWTP showed that total removal of the sum of phthalates in a primary settling tank and SBR was 84% out of which 55% is removed by biodegradation and 29% was removed by sorption to primary and secondary sludge. The percentage removal of four phthalates in primary settling tanks was 18%. Comparison of the diluted effluent DEHP concentration with its environmental quality standards showed that the dilution in an effluent receiving water body can reduce the DEHP emissions to acceptable values.

  2. A review of fast reactor programme in Japan

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  4. Power level control of the TRIGA Mark-II research reactor using the multifeedback layer neural network and the particle swarm optimization

    International Nuclear Information System (INIS)

    Coban, Ramazan

    2014-01-01

    Highlights: • A multifeedback-layer neural network controller is presented for a research reactor. • Off-line learning of the MFLNN is accomplished by the PSO algorithm. • The results revealed that the MFLNN–PSO controller has a remarkable performance. - Abstract: In this paper, an artificial neural network controller is presented using the Multifeedback-Layer Neural Network (MFLNN), which is a recently proposed recurrent neural network, for neutronic power level control of a nuclear research reactor. Off-line learning of the MFLNN is accomplished by the Particle Swarm Optimization (PSO) algorithm. The MFLNN-PSO controller design is based on a nonlinear model of the TRIGA Mark-II research reactor. The learning and the test processes are implemented by means of a computer program at different power levels. The simulation results obtained reveal that the MFLNN-PSO controller has a remarkable performance on the neutronic power level control of the reactor for tracking the step reference power trajectories

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  7. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  10. Power-level regulation and simulation of nonlinear pressurized water reactor core with xenon oscillation using H-infinity loop shaping control

    Directory of Open Access Journals (Sweden)

    Li Gang

    2016-01-01

    Full Text Available This investigation is to solve the power-level control issue of a nonlinear pressurized water reactor core with xenon oscillations. A nonlinear pressurized water reactor core is modeled using the lumped parameter method, and a linear model of the core is then obtained through the small perturbation linearization way. The H∞loop shapingcontrolis utilized to design a robust controller of the linearized core model.The calculated H∞loop shaping controller is applied to the nonlinear core model. The nonlinear core model and the H∞ loop shaping controller build the nonlinear core power-level H∞loop shaping control system.Finally, the nonlinear core power-level H∞loop shaping control system is simulatedconsidering two typical load processes that are a step load maneuver and a ramp load maneuver, and simulation results show that the nonlinear control system is effective.

  11. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  12. Overview of in-vessel retention concept involving level of passivity: with application to evolutionary pressurized water reactor design

    International Nuclear Information System (INIS)

    Ghyym, Seong H.

    1998-01-01

    In this work, one strategy of severe accident management, the applicability of the in-vessel retention (IVR) concept, which has been incorporated in passive type reactor designs, to evolutionary type reactor designs, is examined with emphasis on the method of external reactor vessel cooling (ERVC) to realize the IVR concept in view of two aspects: for the regulatory aspect, it is addressed in the context of the resolution of the issue of corium coolability; for the technical one, the reliance on and the effectiveness of the IVR concept are mentioned. Additionally, for the ERVC method to be better applied to designs of the evolutionary type reactor, the conditions to be met are pointed out in view of the technical aspect. Concerning the issue of corium coolability/quenchability, based on results of the review, plausible alternative strategies are proposed. According to the decision maker's risk behavior, these would help materialize the conceptual design for evolutionary type reactors, especially Korea Next Generation Reactors (KNGRs), which have been developing at the Korea Electric Power Research Institute (KEPRI): (A1) Strategy 1A: strategy based on the global approach using the reliance on the wet cavity method; (A2) Strategy 1B: strategy based on the combined approach using both the reliance on the wet cavity method and the counter-measures for preserving containment integrity; (A3) Strategy 2A: strategy based on the global approach to the reliance on the ERVC method; (A4) Strategy 2B: strategy based on the balanced approach using both the reliance on the ERVC method and the countermeasures for preserving containment integrity. Finally, in application to an advanced pressurized water reactor (PWR) design, several recommendations are made in focusing on both monitoring the status of approaches and preparing countermeasures in regard to the regulatory and the technical aspects

  13. Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brito, Thiago S.P.; Lira, Carlos A.B.O.; Vasconcelos, Wagner E., E-mail: thiago.brito86@yahoo.com.br, E-mail: cabol@ufpe.br, E-mail: wagner@unicap.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociencias. Departamento de Energia Nuclear; Universidade Catolica de Pernambuco (UNICAP), Recife, PE (Brazil). Centro de Ciencias e Tecnologia

    2017-11-01

    It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off). (author)

  14. Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

    International Nuclear Information System (INIS)

    Brito, Thiago S.P.; Lira, Carlos A.B.O.; Vasconcelos, Wagner E.; Universidade Catolica de Pernambuco

    2017-01-01

    It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off). (author)

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

    International Nuclear Information System (INIS)

    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)

  16. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  17. Investigation of free level fluctuations in a simulated model of a sodium cooled Fast Breeder Reactor using pulsating conductance monitoring device

    International Nuclear Information System (INIS)

    Aggarwal, P.K.; Pandey, G.K.; Malathi, N.; Arun, A.D.; Ananthanarayanan, R.; Banerjee, I.; Sahoo, P.; Padmakumar, G.; Murali, N.

    2012-01-01

    Highlights: ► An innovative approach for measurement of water level fluctuation is presented. ► Measurement was conducted with a PC based pulsating type level sensor. ► Deployed the technique in monitoring level fluctuation in PFBR simulated facility. ► The technique helped in validation of hot pool design of PFBR, India. - Abstract: A high resolution measurement technique for rapid and accurate monitoring of water level using an in-house built pulsating conductance monitoring device is presented. The technique has the capability of online monitoring of any sudden shift in water level in a reservoir which is subjected to rapid fluctuations due to any external factor. We have deployed this novel technique for real time monitoring of water level fluctuations in a specially designed ¼ scale model of the Prototype Fast Breeder Reactor (PFBR) at Kalpakkam, India. The water level measurements in various locations of the simulated test facility were carried out in different experimental campaigns with and without inclusion of thermal baffles to it in specific operating conditions as required by the reactor designers. The amplitudes and the frequencies of fluctuations with required statistical parameters in hot water pool of the simulated model were evaluated from the online time versus water level plot in more convenient way using system software package. From experimental results it is computed that the maximum free level fluctuation in the hot pool of PFBR with baffle plates provided on the inner vessel is 30 mm which is considerably less than the value (∼82 mm) obtained without having any baffle plates. The present work provided useful information for assessment of appropriate design which would be adopted in the PFBR for safe operation of the reactor.

  18. Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

    2010-03-01

    The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the probabilistic safety assessment (PSA) domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

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

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    2012-01-01

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

  20. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  1. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  2. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

    Mangus, J.D.; Cooper, M.H.

    1982-01-01

    An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

  3. Lessons Learnt in the Development of Level 1 PUSPATI TRIGA Reactor Probability Safety Assessment: A Collaboration Project under the Norwegian Extra Budgetary Fund

    International Nuclear Information System (INIS)

    Mazleha Maskin; Tom, P.P.; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Muhamad Puad Abu

    2014-01-01

    This article reports about the lessons learnt from the development of level 1 probabilistic safety assessment (PSA) project that was implemented under the IAEA mentoring program for TRIGA MARK II PUSPATI research reactor (RTP). As a project that involved more than 3 organizations, a strategic planning of the management and implementation of individual assignment is truly a hectic task. This report compiles all related activities from the forming of the Malaysian PSA team up to the final report submitted to the IAEA. (author)

  4. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  5. Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

    Energy Technology Data Exchange (ETDEWEB)

    Johanna Oxstrand; Katya Le Blanc

    2014-07-01

    The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

  6. Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

    2015-06-19

    This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

  7. Estimative of core damage frequency in IPEN's IEA-R1 research reactor (PSA level 1) due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami

    2009-01-01

    This work applies the methodology of probabilistic safety assessment level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid by major pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, emergency core cooling system (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  8. Determination of reactor parameters in a simulated RA reactor lattice by measuring the reactivity level of heavy water in D{sub 2}O moderated RB reactor; Odredjivanje reaktorskih parametara u simuliranoj resetki reaktora RA merenjem reaktivnosti nivoa teske vode u D{sub 2}O moderiranom reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Takac, S M; Markovic, H D; Dimitrijevic, Z B [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1974-07-01

    Direct method for experimental determination of the neutron age {tau} in the reactor lattice is not developed. Fermi theory for determining {tau} by measuring the reactivity level of the heavy water can be applied for a limited number of reactor lattices. An attempt was made to apply this approach for a complex reactor core with side and upper reflector. As expected the obtained results were not satisfactory, and {tau} was determined by Dessauer formula which gives more realistic estimation of thermalization in the reactor cell. major discrepancies are resulting from the fact that the lower reflector was neglected. But it is possible to to determine {tau} for reactor core with reflectors by additional measurements of axial distribution and other experimental data for {tau}. This is quite tedious numerical procedure. Obtained experimental data for a number of reactor parameters are compared to the initial data of the RA reactor core showing very good agreement. [Serbo-Croat] Ekspeimentalna metoda za direktno odredjivanje starosti neutrona {tau}, u resetki reaktora do danas nije razvijena. Primenjena Fermijeova teorija za odredjivanje {tau} preko merenja reaktivnosti nivoa teske vode, ocigledno se moze koristiti samo na ogranicen spektar reaktorskih resetki. U ovom radu pokusano je da se vidi primenljivost iste u slozenom - bocno i odozdo - reflektovanom jezgru reaktora. Naravno dobijeni rezultati nisu zadovoljili, sto se i moglo ocekivati, pa je {tau} odredjen preko Dessauer-ove formule, sto daje daleko blizu sliku stvarnog stanja procesa termalizacije u celiji reaktora. Ocigledno vece neslaganje primenjene teorije dolazi od zanemarivanja donjeg reflektora u jezgru reaktora. Medjutim, dopnskim merenjem aksijalne raspodele i na osnovu ostalih eksperimentalnih podataka za {tau}, moguce je odrediti eksperimentalno reflektorski koeficijent za visestruko reflektovano jezgro reaktora, sto numericki predstavlja mukotrpan i dugotrajan rad na racunaru. Dobijeni

  9. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

    International Nuclear Information System (INIS)

    Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

    2009-01-01

    The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

  10. The influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low-alloy steels under simulated boiling water reactor conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.

    2016-01-01

    Highlights: • Chloride effects on SCC crack growth in RPV steels under boiling water reactor conditions. • ppb-levels of chloride may result in fast SCC in normal water chemistry environment. • Much higher chloride tolerance for SCC in hydrogen water chemistry environment. • Potential long-term (memory) effects after severe and prolonged temporary chloride transients. - Abstract: The effect of chloride on the stress corrosion crack (SCC) growth behaviour in low-alloy reactor pressure vessel steels was evaluated under simulated boiling water reactor conditions. In normal water chemistry environment, ppb-levels of chloride may result in fast SCC after rather short incubation periods of few hours. After moderate and short-term chloride transients, the SCC crack growth rates return to the same very low high-purity water values within few 100 h. Potential long-term (memory) effects on SCC crack growth cannot be excluded after severe and prolonged chloride transients. The chloride tolerance for SCC in hydrogen water chemistry environment is much higher.

  11. Determination of maximum reactor power level consistent with the requirement that flow reversal occurs without fuel damage

    International Nuclear Information System (INIS)

    Rao, D.V.; Darby, J.L.; Ross, S.B.; Clark, R.A.

    1990-01-01

    The High Flux Beam Reactor (HFBR) operated by Brookhaven National Laboratory (BNL) employs forced downflow for heat removal during normal operation. In the event of total loss of forced flow, the reactor will shutdown and the flow reversal valves open. When the downward core flow becomes sufficiently small then the opposing thermal buoyancy induces flow reversal leading to decay heat removal by natural convection. There is some uncertainty as to whether the natural circulation is adequate for decay heat removal after 60 MW operation. BNL- staff carried out a series of calculations to establish the adequacy of flow reversal to remove decay heat. Their calculations are based on a natural convective CHF model. The primary purpose of the present calculations is to review the accuracy and applicability of Fauske's CHF model for the HFBR, and the assumptions and methodology employed by BNL-staff to determine the heat removal limit in the HFBR during a flow reversal and natural convection situation

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Exercising Synergy of Safeguards Safety and Security at Facility Level of the GA Siwabessy Multi-Purpose Reactor, Indonesia

    International Nuclear Information System (INIS)

    Susilowati, E.

    2015-01-01

    Safeguards, safety and security (3Ss) constitute as essential elements for successful development of nuclear technology in the life time of nuclear installation. All 3Ss need to be coordinated due workers, the public and the environment require protection from plant malfunction, human error, malicious acts and proliferation of nuclear materials and technologies. Then the importance of the 3Ss was deemed valuable, particularly to a country having willingness to expand to nuclear power reactor such as Indonesia that in the near future plans to build small experimental power reactor. This paper is aimed to discuss synergy among safeguards, safety and security which will have opportunity been exercising at the GA Siwabessy Reactor (RSG-GAS), Indonesia. Synergy among safeguards, safety and security offers much opportunity for cost savings and enhance efficiency. Discussion is carried out by first investigating common values and conflicts exist among 3S. Up to now each of them was accomplished separately by different division and using different equipment due lack of coordination among them. The objective of this exercise is to develop more efficient and effective 3Ss infrastructures and also to support skill and knowledge of human resources. Benefitting from synergy between safeguards and security such as management of nuclear material and non proliferation; safeguards and safety such as management of nuclear material and waste management; safety and security such as prevent radiological release and also tension among them if any are discussed. It is expected that outcome of this exercise will able to develop a role model of infrastructures to the up-coming small experimental power reactor in Indonesia. (author)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  18. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  19. Collection and transmission of milli-volt-level data on a liquid-metal fast-breeder reactor

    International Nuclear Information System (INIS)

    Robinson, C.; Thoburn, R.

    1978-01-01

    The paper describes a high integrity data multiplexing system which caters for up to 252 input elements with remote measuring facilities, using a 12-core flexible cable. Tests carried out on a prototype unit and recommendations for effecting improvements are described. The system was developed following the identification of the problem of having to handle a large number of cables on the top of the Commercial Demonstration Fast Reactor. It has high reliability, a relatively small and conventional cable installation and is applicable to situations outside the nuclear field. (author)

  20. Detailed description of an SSAC at the facility level for on-load refueled power reactor facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-11-01

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in an on-load refueled power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  3. A Review of the Fast Reactor Programme in Japan

    International Nuclear Information System (INIS)

    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)

  4. Integral test of JENDL-3.3 on fast reactors

    International Nuclear Information System (INIS)

    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)

  5. A review and analysis of European industrial experience in handling LWR [light water reactor] spent fuel and vitrified high-level waste

    International Nuclear Information System (INIS)

    Blomeke, J.O.

    1988-06-01

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performances of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States. 79 refs., 71 figs., 10 tabs

  6. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    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

  7. Reactor Operations informal monthly report December 1994

    International Nuclear Information System (INIS)

    1994-12-01

    Reactor operations at the MRR and HFBR reactors at Brookhaven National Laboratory are presented for December 1994. Reactor run-time and power levels, instrumentation, mechanical maintenance, occurrence reports, and safety information are included

  8. Detailed description of an SSAC at the facility level for light water moderated (off-load refueled) power reactor facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-03-01

    This report is intended to provide the technical details of an effective State Systems of Accounting for and Control of Nuclear Material (SSAC) which Member States may use, if they wish, to establish and maintain their SSACs. It is expected that systems designed along the lines described would be effective in meeting the objectives of both national and international systems for nuclear material accounting and control. This document accordingly provides a detailed description of a system for the accounting for and control of nuclear material in an off-load refueled light water moderated power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  9. Greater-than-Class C low-level radioactive waste characterization. Appendix A-3: Basis for greater-than-Class C low-level radioactive waste light water reactor projections

    International Nuclear Information System (INIS)

    Mancini, A.; Tuite, P.; Tuite, K.; Woodberry, S.

    1994-09-01

    This study characterizes low-level radioactive waste types that may exceed Class C limits at light water reactors, estimates the amounts of waste generated, and estimates radionuclide content and distribution within the waste. Waste types that may exceed Class C limits include metal components that become activated during operations, process wastes such as cartridge filters and decontamination resins, and activated metals from decommissioning activities. Operating parameters and current management practices at operating plants are reviewed and used to estimate the amounts of low-level waste exceeding Class C limits that is generated per fuel cycle, including amounts of routinely generated activated metal components and process waste. Radionuclide content is calculated for specific activated metals components. Empirical data from actual low-level radioactive waste are used to estimate radionuclide content for process wastes. Volumes and activities are also estimated for decommissioning activated metals that exceed Class C limits. To estimate activation levels of decommissioning waste, six typical light water reactors are modeled and analyzed. This study does not consider concentration averaging

  10. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  11. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  12. Reactor feedwater control device

    International Nuclear Information System (INIS)

    Koshi, Yuji.

    1993-01-01

    In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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.)

  15. X-ray radiographic experimental investigation of the reference level in hydrostatic level measurement systems for boiling water reactors; Roentgen-radiografische Untersuchung des Referenzfuellstandes bei der hydrostatischen Fuellstandsmessung in Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, S.; Hampel, R. [Hochschule Zittau/Goerlitz, Zittau (Germany). Inst. fuer Prozesstechnik, Prozessautomatisierung und Messtechnik (IPM); Boden, S. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Dresden (Germany). Inst. fuer Fluiddynamik

    2012-11-01

    Hydrostatic level control is of high priority for normal operation safety of BWR-type reactors. For the prevention of undershooting the limiting values precise and secure measuring techniques are indispensable. Actually the filling level is determined form the hydrostatic pressure difference to a reference column. For the secure non-invasive detection of the phase boundary water/steam in inclined tubes the X-ray radiography has been chosen. The experiments were aimed to study possible geometric influences on the water/steam phase boundary. It was shown that the reference filling level is not significantly changed in spite of permanent phase transitions, provided an ideal mechanical construction of the system is given. Future experiments shall be focused on the analysis of interface behavior in case of non-ideal geometries (welds).

  16. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  17. Australian research reactor studies

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1978-01-01

    The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

  18. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  19. Development of nuclear thermal hydraulic verification tests and evaluation technology - Development of the ultrasonic method for two-phase mixture level measurement in nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Kim, Sang Jae; Kim, Hyung Tae; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    An ultrasonic method is developed for the measurement of the two-phase mixture level in the reactor vessel or steam generator. The ultrasonic method is selected among the several non-nuclear two-phase mixture level measurement methods through two steps of selection procedure. A commercial ultrasonic level measurement method is modified for application into the high temperature, pressure, and other conditions. The calculation method of the ultrasonic velocity is modified to consider the medium as the homogeneous mixture of air and steam, and to be applied into the high temperature and pressure conditions. The cross-correlation technique is adopted as a detection method to reduced the effects of the attenuation and the diffused reflection caused by surface fluctuation. The waveguides are developed to reduce the loss of echo and to remove the effects of obstructs. The present experimental study shows that the developed ultrasonic method measures the two-phase mixture level more accurately than the conventional methods do. 21 refs., 60 figs., 13 tabs. (Author)

  20. Measurement and regulation of the level of a homogeneous plutonium reactor; Mesure et regulation du niveau d'un reacteur homogene au plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Berger, F; Bertrand, J

    1958-12-01

    Reactivity depends strongly on disturbances of the level of the plutonium solution In the homogeneous reactor. Proserpine has a small cylindrical core, 250 mm diameter, and 10 liters volume. With a view to reducing the dangers due to corrosion and contamination, the solution level in the core is raised by pneumatic pressure. The level is stabilized by means of a regulating system. During critical experiments the variations of the level are less than one hundredth part of a millimeter. (author) [French] Les variations du niveau de la solution de plutonium dans le reacteur homogene Proserpine ont une grosse influence sur la reactivite, car le coeur est petit (10 litres de solution dans un cylindre de diametre 250 mm). En vue de reduire les dangers dus a la corrosion et a la contamination, la commande du volume liquide est pneumatique. Nous avons realise la stabilite du niveau par une regulation qui, dans les essais en regime critique, limite les variations du plan liquide a une fraction de centieme de millimetre. (auteur)

  1. Photocatalytic performance of cylindrical reactor inserted with UV light-emitting-diodes for purification of low-level toxic volatile organic compounds

    International Nuclear Information System (INIS)

    Jo, Wan K.; Kang, Hyun J.

    2012-01-01

    Highlights: ► Photocatalyst baked at 350 °C exhibited the highest BTEX degradation efficiency. ► Conventional lamp evidenced a higher degradation efficiency compared to LEDs. ► LEDs was more energy-efficient than conventional lamp for BTEX degradation. ► As the residence time increased, the average degradation efficiency increased. - Abstract: The present study investigated the photocatalytic performance of a cylindrical reactor inserted with UV light-emitting-diodes for the decomposition of low-level (0.1 ppm) gas-phase organic compounds (benzene, toluene, ethyl benzene and xylene (BTEX)). The morphological and optical properties of photocatalysts (Degussa P-25 TiO 2 ) baked at different temperatures were determined using a range of spectral instruments. The photocatalyst baked at 350 °C exhibited the highest conversion efficiencies for both benzene and toluene (81 and ∼100%, respectively). The conventional lamp showed a higher conversion efficiency for benzene compared to the 380-nm UV-LED and a higher conversion efficiency for benzene and toluene than the 365-nm UV-LED. However, the ratios of conversion efficiency to electric power consumption were 2.5–3.0 times higher for the latter light source than the former source. Moreover, as the residence time increased from 0.2 to 1.2 min, the average conversion efficiencies for BTEX of the 3-h photocatalytic process increased from nearly zero to 81%, 7 to nearly 100%, 20 to nearly 100%, and 29–30 to nearly 100%, respectively. The cylindrical photocatalytic reactor inserted with UV-LEDs could be energy-efficiently applied for the decomposition of low-level toxic compounds after optimization of the operating conditions.

  2. Photocatalytic performance of cylindrical reactor inserted with UV light-emitting-diodes for purification of low-level toxic volatile organic compounds

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Wan K., E-mail: wkjo@knu.ac.kr [Department of Environmental Engineering, Kyungpook National University, Daegu, 702-701 (Korea, Republic of); Kang, Hyun J., E-mail: khj435@naver.com [Department of Environmental Engineering, Kyungpook National University, Daegu, 702-701 (Korea, Republic of)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Photocatalyst baked at 350 Degree-Sign C exhibited the highest BTEX degradation efficiency. Black-Right-Pointing-Pointer Conventional lamp evidenced a higher degradation efficiency compared to LEDs. Black-Right-Pointing-Pointer LEDs was more energy-efficient than conventional lamp for BTEX degradation. Black-Right-Pointing-Pointer As the residence time increased, the average degradation efficiency increased. - Abstract: The present study investigated the photocatalytic performance of a cylindrical reactor inserted with UV light-emitting-diodes for the decomposition of low-level (0.1 ppm) gas-phase organic compounds (benzene, toluene, ethyl benzene and xylene (BTEX)). The morphological and optical properties of photocatalysts (Degussa P-25 TiO{sub 2}) baked at different temperatures were determined using a range of spectral instruments. The photocatalyst baked at 350 Degree-Sign C exhibited the highest conversion efficiencies for both benzene and toluene (81 and {approx}100%, respectively). The conventional lamp showed a higher conversion efficiency for benzene compared to the 380-nm UV-LED and a higher conversion efficiency for benzene and toluene than the 365-nm UV-LED. However, the ratios of conversion efficiency to electric power consumption were 2.5-3.0 times higher for the latter light source than the former source. Moreover, as the residence time increased from 0.2 to 1.2 min, the average conversion efficiencies for BTEX of the 3-h photocatalytic process increased from nearly zero to 81%, 7 to nearly 100%, 20 to nearly 100%, and 29-30 to nearly 100%, respectively. The cylindrical photocatalytic reactor inserted with UV-LEDs could be energy-efficiently applied for the decomposition of low-level toxic compounds after optimization of the operating conditions.

  3. Management of solid wastes during decommissioning of research reactors. Evaluation of gross clearance levels and mathematical simulation of solid waste assay techniques

    International Nuclear Information System (INIS)

    Gopalakrishnan, R.K.; Sobhan Babu, K.; Sharma, D.N.

    2008-01-01

    Full text: Decommissioning of nuclear facilities constitute a challenge mainly due to the huge and complex nature of radioactive waste generated during this process. In the context of management and disposal of waste and reuse/recycle of usable materials during decommissioning of reactors, clearance levels for relevant radio nuclides are of vital importance. Radionuclide specific clearance levels are developed by IAEA and such levels allow the facility for free release of materials to the environment without further regulatory consideration. An effort has been made in this paper to establish clearance levels for radionuclides associated with various system and structural components of a research reactor and rather than radionuclide specific clearance levels, these values are derived for gross activity concentration, which is more practical for radioactive waste categorization, disposal and reuse or recycle of usable materials. The first step towards the derivation of clearance levels is the calculation of annual doses relating to unit activity concentration for each nuclide using various enveloping scenarios. After the estimation of doses, the limiting enveloping scenario (the one that gives the highest dose) is identified. The clearance levels are then derived by dividing the reference dose level (10 μSv/y) by the annual dose calculated per unit activity concentration for the limiting enveloping scenario The clearance level for gross beta-gamma activity concentration is then evaluated as the product of the limiting clearance level and the number of radionuclides characterized for the structural components. Simulation studies were also carried out for the design of a monitoring system for estimation of activity concentration of the decommissioned materials, especially rubbles/ concrete, using mathematical models. Conventional solid waste assay techniques would not suffice to the requirement of decommissioning waste categorization since very low level activity

  4. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

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

  5. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  6. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  7. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  8. Multi-purpose reactor

    International Nuclear Information System (INIS)

    1991-05-01

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

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

    International Nuclear Information System (INIS)

    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.)

  10. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    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

  13. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  14. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  15. Low-level radioactive waste from commercial nuclear reactors. Volume 4. Proceedings of the workshop on research and development needs for treatment of low-level radioactive waste from commercial nuclear reactors

    International Nuclear Information System (INIS)

    Godbee, H.W.; Frederick, E.J.; Jolley, R.L.; Kibbey, A.H.; Rodgers, B.R.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. As part of this program, a workshop was conducted for determining research and development needs in LLRW treatment. Volume 4, the proceedings of this workshop, includes the formal presentations and both panel and general discussions dealing with such issues as disposal, compaction, and the ''below regulatory concern'' philosophy. Summaries of individual workshops dealing with specific aspects of LLRW treatment are also presented in this volume

  16. Low-level radioactive waste from commercial nuclear reactors. Volume 1. Recommendations for technology developments with potential to significantly improve low-level radioactive waste management

    International Nuclear Information System (INIS)

    Rodgers, B.R.; Jolley, R.L.

    1986-02-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 1 provides an executive summary and a general introduction to the four-volume set, in addition to recommendations for research and development (R and D) for low-level radioactive waste (LLRW) treatment. Generic, long-range, and/or high-risk programs identified and prioritized as needed R and D in the LLRW field include: (1) systems analysis to develop decision methodology; (2) alternative processes for dismantling, decontaminating, and decommissioning; (3) ion exchange; (4) incinerator technology; (5) disposal technology; (6) demonstration of advanced technologies; (7) technical assistance; (8) below regulatory concern materials; (9) mechanical treatment techniques; (10) monitoring and analysis procedures; (11) radical process improvements; (12) physical, chemical, thermal, and biological processes; (13) fundamental chemistry; (14) interim storage; (15) modeling; and (16) information transfer. The several areas are discussed in detail

  17. Reactor utilization, Part 1

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1981-01-01

    The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  19. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  20. Determination of the exposition rapidity in the level 49.90 of the reactor building for the decrease in the water level of the spent fuel pool; Determinacion de la rapidez de exposion en el nivel 49.90 del edificio del reactor por la disminucion en el nivel de agua de la alberca de combustible gastado

    Energy Technology Data Exchange (ETDEWEB)

    Mijangos D, Z. E.; Herrera H, S. F.; Cruz G, M. A.; Amador C, C., E-mail: zoedelfin@gmail.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The fuel assemblies storage in the nuclear power plant of Laguna Verde (NPP-L V) represents a crucial aspect, due to the generated dose by the decay heat of the present radio-nuclides in the assemblies retired of the reactor core, after their useful life. These spent assemblies are located inside the spent fuel pool (SFP), in the level 49.90 m in the Reload Floor of the Reactor building of NPP-L V. This leads to the protection at personnel applying the ALARA (As Low As Reasonably Achievable) criteria, fulfilling the established dose criteria by the Regulator Body the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). Considering the loss scenario of the cooling system of the SFP, in which the SFP water vaporizes, is important to know the water level in which the limit of effective dose equivalent is fulfilled for the personnel. Also, is important for the instrumentation of the SFP, for the useful life of the same instruments. In this work is obtained the exposition rapidity corresponding to different water levels of SFP in the Reload Floor of NPP-L V, to identify the minimum level of water where the limit of effective dose equivalent is fulfilled of 25 rem s to the personnel, established in the Article 48 of the General Regulation of Radiological Safety of CNSNS and the Chapter 50 Section 67 of the 10-Cfr of Nuclear Regulatory Commission in USA. The water level is also identified where the exposition rapidity is of 15 m R/hr, being the value of the set point of the area radiation monitor D21-Re-N003-1, located to 125 cm over the level 49.90 meters of the Reload Floor of NPP-L V. (Author)

  1. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  2. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  3. Low-level radioactive waste from commercial nuclear reactors. Volume 2. Treatment, storage, disposal, and transportation technologies and constraints

    Energy Technology Data Exchange (ETDEWEB)

    Jolley, R.L.; Dole, L.R.; Godbee, H.W.; Kibbey, A.H.; Oyen, L.C.; Robinson, S.M.; Rodgers, B.R.; Tucker, R.F. Jr.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 2 discusses the definition, forms, and sources of LLRW; regulatory constraints affecting treatment, storage, transportation, and disposal; current technologies used for treatment, packaging, storage, transportation, and disposal; and the development of a matrix relating treatment technology to the LLRW stream as an aid for choosing methods for treating the waste. Detailed discussions are presented for most LLRW treatment methods, such as aqueous processes (e.g., filtration, ion exchange); dewatering (e.g., evaporation, centrifugation); sorting/segregation; mechanical treatment (e.g., shredding, baling, compaction); thermal processes (e.g., incineration, vitrification); solidification (e.g., cement, asphalt); and biological treatment.

  4. Low-level radioactive waste from commercial nuclear reactors. Volume 2. Treatment, storage, disposal, and transportation technologies and constraints

    International Nuclear Information System (INIS)

    Jolley, R.L.; Dole, L.R.; Godbee, H.W.; Kibbey, A.H.; Oyen, L.C.; Robinson, S.M.; Rodgers, B.R.; Tucker, R.F. Jr.

    1986-05-01

    The overall task of this program was to provide an assessment of currently available technology for treating commercial low-level radioactive waste (LLRW), to initiate development of a methodology for choosing one technology for a given application, and to identify research needed to improve current treatment techniques and decision methodology. The resulting report is issued in four volumes. Volume 2 discusses the definition, forms, and sources of LLRW; regulatory constraints affecting treatment, storage, transportation, and disposal; current technologies used for treatment, packaging, storage, transportation, and disposal; and the development of a matrix relating treatment technology to the LLRW stream as an aid for choosing methods for treating the waste. Detailed discussions are presented for most LLRW treatment methods, such as aqueous processes (e.g., filtration, ion exchange); dewatering (e.g., evaporation, centrifugation); sorting/segregation; mechanical treatment (e.g., shredding, baling, compaction); thermal processes (e.g., incineration, vitrification); solidification (e.g., cement, asphalt); and biological treatment

  5. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  7. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V.; Galvin, M.; Todreas, N.E.; Lombardi, C.V.; Maldari, F.; Ricotti, M.E.; Cinotti, L.

    2001-01-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  8. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  9. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  10. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  11. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  12. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

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

    International Nuclear Information System (INIS)

    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.

  14. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  15. Reactor protecting device

    International Nuclear Information System (INIS)

    Ono, Hiroshi; Kasuga, Hajime; Kasuga, Hiroshi.

    1984-01-01

    Purpose: To reduce the recycling flowrate thereby decrease the neutron flux level before the reactor shutdown upon generation of abnormality such as increase in the neutron flux, by setting the safety level lower than the value for generating the reaction scram signal. Constitution: A netron flux safety level setter and an instruction signal generator are disposed between a neutron flux detector and a recycling flowrate control device. A neutron flux safety level lower than the level for generating a reactor scram signal and higher that the level for the ordinary operation is set and, if the detection level for the neutron flux in the reactor core arrives at the safety level, a neutron flux decreasing instruction signal is outputted from the instruction signal generator to the recycling flowrate control device to thereby decrease the recycling flowrate and decrease the neutron flux without reaching the reactor shutdown, whereby the thermal safety of the fuel rod can be maintained and the reactor operation performance can be improved. (Moriyama, K.)

  16. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    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

  17. Feed water control device in a reactor

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1984-01-01

    Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

  18. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  19. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  20. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  1. Photocatalytic degradation of mixed gaseous carbonyl compounds at low level on adsorptive TiO2/SiO2 photocatalyst using a fluidized bed reactor.

    Science.gov (United States)

    Zhang, Maolin; An, Taicheng; Fu, Jiamo; Sheng, Guoying; Wang, Xinming; Hu, Xiaohong; Ding, Xuejun

    2006-06-01

    An adsorptive silica-supported titania photocatalyst TiO(2)/SiO(2) was prepared by using nanosized titania (anatase) immobilized on silica gel by the sol-gel technique with the titanium tetra isopropoxide as the main raw material and acetic acid as the acid catalyst. Meanwhile the structure and properties of the TiO(2)/SiO(2) photocatalyst were studied by means of many modern analysis techniques such as TEM, XRD, and BET. Gas-solid heterogeneous photocatalytic decomposition of four carbonyl compounds mixture at low concentration levels over ultraviolet irradiated TiO(2)/SiO(2) photocatalyst were carried out with high degradation efficiencies in a coaxial triple-cylinder-type fluidized bed photocatalytic reactor, which provided efficient continuous contact of ultraviolet photons, silica-supported titania photocatalyst, and gaseous reactants. Experimental results showed that the photocatalyst had a high adsorption performance and a good photocatalytic activity for four carbonyl compounds mixture. Some factors influencing the photocatalytic decomposition of the mixed carbonyl compounds, i.e. the gas flowrate, relative humidity, concentration of oxygen, and illumination time, were discussed in detail. It is found that the photocatalytic reaction rate of four carbonyl compounds decreased in this order: propionaldehyde, acetone, acetaldehyde and formaldehyde.

  2. Evaluation of the Impact That PARs Have on the Hydrogen Risk in the Reactor Containment: Methodology and Application to PSA Level 2

    Directory of Open Access Journals (Sweden)

    Ahmed Bentaib

    2010-01-01

    Full Text Available This paper presents a methodology and its application to a Level 2 Probabilistic Safety Assessment (PSA-2, to evaluate the impact of the Passive Autocatalytic Recombiners (PARs on the hydrogen risk in the reactor containment in case of a severe accident. Among the whole set of accidental scenarios calculated in the framework of the PSA-2, nine have been selected as representative in terms of the in-vessel hydrogen production rate and in-vessel total produced hydrogen mass. Five complementary scenarios have been added as representative of the core reflooding situations. For this set of selected scenarios the evolution of the conditions in the containment (i.e., pressure, temperature, and composition during the in-vessel phase of the accident has been evaluated by means of a lumped parameter approach. The use of spray systems in the containment has also been considered as well as the presence of recombiners. Moreover, the ignition by recombiners of the flammable atmosphere has been considered.

  3. Reports in the area sensor technology: Part 2: Dynamic deviations which may appear via the sensing lines in measurement of reactor pressure and level

    International Nuclear Information System (INIS)

    Bergdahl, Bengt-Goeran

    2002-12-01

    demand on the response time was fulfilled. The report presents how blocking of the sensing lines can be supervised. Examples are given from the Muehleberg NPP (KKM). Examples are shown with a flow measurement transmitter, which has pulsation dampers. For this presentation, results with GSE's database for sensor tests, SensBase, are used. SensBase stores new results every year. In this way, the influence of component ageing can also be detected. Because of the use of SensBase, the Swiss nuclear power inspectorate has approved less frequent calibration of the transmitters. The report also presents pressure oscillations that take place in the sensing line and not in the real measured process. The water in the sensing line together with the transmitter membrane form a dynamic system with water as mass, elasticity in the transmitter membrane as spring constant and reactor pressure fluctuations as driving force. The problem with oscillations in the measurement system is illustrated with examples from Ringhals 1 and KKM. Another example from KKM shows how the oscillation in a level transmitter of the type Barton Cell affected a large number of transmitters, which were all connected to the common sensing lines. It was possible to identify the deviating transmitter by experiments with isolation valve closures in the sensing lines during power operation. After the transmitter had been changed to one with a smaller volume and smaller displacement, the oscillations disappeared. Finally, the report also shows that mechanical vibrations in the sensing lines contribute to the signal noise around 10 Hz. This is shown by means of laboratory tests at KKM. At this plant, transmitters have also been replaced because of deviating noise in the frequency interval 2-20 Hz. After the transmitter had been replaced, the noise disappeared. The results from KKM show that it can never be excluded that ageing may increase the transmitter's sensitivity to vibrations of the sensing line

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

    International Nuclear Information System (INIS)

    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

  5. Remote level radiation monitoring system for the brazilian IEA-R1 nuclear research reactor for routine radiation protection procedures and as a support tool in case of radiological emergency

    International Nuclear Information System (INIS)

    Cardenas, Jose P.N.; Romero Filho, Christovam R.; Madi Filho, Tufic

    2008-01-01

    Nuclear facilities must monitoring radiation levels to establish procedures for radiological protection staff involving workers and the public. The Instituto de Pesquisas Energeticas e Nucleares - IPEN has 5 important plants and in case of accident in one of them, the Institute keeps operational an Emergency Response Plan (ERP). This document (ERP) is designed to coordinate all procedures to assure safe and secure conditions for workers, environment and the public. One of this plants is the IEA-R1 reactor, it is the oldest nuclear research reactor (pool type) in Latin America, reached it first criticality in September of 1957. The reactor is used 60 hours/week with continuous operation and with nominal power of 3.5 MW, with technical conditions to operate at 5 MW thermal power. This reactor has a Radiological Emergency Plan that establishes the implementation of rules for workers and people living at the exclusion area in the case of an emergency situation. This paper aims to describe the implementation of a computational system developed for remote radiation monitoring, in a continuous schedule of IEA-R1 nuclear research reactor containment building. Results of this action can be used as a support mean in a radiological emergency. All necessary modules for radiation detection, signals conditioners and processing, data acquisition board, software development and computer specifications are described. The data acquisition system operating in the reactor shows readings concerned to radiation environment such as activity, doses and concentration in real time and displays a periodical data bank (Data Base) of this features allowing through the surveillance of the operation records anytime, leading to studies and analysis of radiation levels. Results of this data acquisition are shown by means of computer graphics screens developed for windows environment using Visual Basic software. (author)

  6. Determination of scaling factors to estimate the radionuclide inventory in waste with low and intermediate-level activity from the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Taddei, Maria Helena Tirollo

    2013-01-01

    Regulations regarding transfer and final disposal of radioactive waste require that the inventory of radionuclides for each container enclosing such waste must be estimated and declared. The regulatory limits are established as a function of the annual radiation doses that members of the public could be exposed to from the radioactive waste repository, which mainly depend on the activity concentration of radionuclides, given in Bq/g, found in each waste container. Most of the radionuclides that emit gamma-rays can have their activity concentrations determined straightforwardly by measurements carried out externally to the containers. However, radionuclides that emit exclusively alpha or beta particles, as well as gamma-rays or X-rays with low energy and low absolute emission intensity, or whose activity is very low among the radioactive waste, are generically designated as Difficult to Measure Nuclides (DTMs). The activity concentrations of these DTMs are determined by means of complex radiochemical procedures that involve isolating the chemical species being studied from the interference in the waste matrix. Moreover, samples must be collected from each container in order to perform the analyses inherent to the radiochemical procedures, which exposes operators to high levels of radiation and is very costly because of the large number of radioactive waste containers that need to be characterized at a nuclear facility. An alternative methodology to approach this problem consists in obtaining empirical correlations between some radionuclides that can be measured directly – such as 60 Co and 137 Cs, therefore designated as Key Nuclides (KNs) – and the DTMs. This methodology, denominated Scaling Factor, was applied in the scope of the present work in order to obtain Scaling Factors or Correlation Functions for the most important radioactive wastes with low and intermediate-activity level from the IEA-R1 nuclear research reactor. (author)

  7. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  8. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

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

    International Nuclear Information System (INIS)

    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. RB Research nuclear reactor, Annual report for 2005

    International Nuclear Information System (INIS)

    Milosevic, M.; Dasic, N.; Ljubenov, V.; Pesic, M.; Nikolic, D; Jevremovic, M.; Minic, D.

    2006-01-01

    Report on RB reactor operation during 2005 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2005

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

    International Nuclear Information System (INIS)

    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

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  15. Physical security at research reactors

    International Nuclear Information System (INIS)

    Clark, R.A.

    1977-01-01

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  16. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  17. Test reactors in the world

    International Nuclear Information System (INIS)

    Corella, M.R.; Gomez Alonso, M.

    1983-01-01

    INFCE work on research reactor core conversion from HEU to LEU, attracted a raising interest on this type of nuclear reactors. In this context, the present work shows a compilation of worldwide research and test nuclear reactors, now in operation, under construction, or planned, as well as decommissioned reactors (tables A to F). Brief descriptions of these reactors are included in tables G to L. In table M a summary view of reactors with power level between 10 and 30 MWt is shown. Attention is focused on that power range, as it has been considered in very preliminar studies for a new research reactor. Almost all data have been obtained from current available bibliography. (author)

  18. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  19. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  20. Revitalization of reactor usage through reactor sharing

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1986-01-01

    The purpose of this work was to renew interest in using the University Florida Training Reactor (UFTR) for educational and training purposes outside the Nuclear Engineering Sciences (NES) and Environmental Engineering Sciences (EES) Departments at the University of Florida and for research by others outside the NES Department. The availability of the UFTR made possible through a US Department of Energy (DOE) Reactor Sharing Grant provided the mechanism to pursue generation of renewed interest at all levels both within the University of Florida and from other educational and corporate institutions

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  3. Initiating Events for Multi-Reactor Plant Sites

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

  4. Fast breeder reactors

    International Nuclear Information System (INIS)

    Waltar, A.E.; Reynolds, A.B.

    1981-01-01

    This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of reactor theory, heat transfer, and fluid mechanics. The book is expected to be used most widely for a one-semester general course on fast breeder reactors, with the extent of material covered to vary according to the interest of the instructor. The book could also be used effectively for a two-quarter or a two-semester course. In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety. Methodology in fast reactor design and analysis, together with physical descriptions of systems, is emphasized in this text more than numerical results. Analytical and design results continue to change with the ongoing evolution of FBR design whereas many design methods have remained fundamentally unchanged for a considerable time

  5. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    Gubel, P.

    2001-01-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  6. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  7. Complement of existing ASAMPSA2 guidance for Level 2 PSA for shutdown states of reactors, Spent Fuel Pool and recent R and D results

    International Nuclear Information System (INIS)

    Kumar, M.; Olsson, A.; Loeffler, H.; Morandi, S.; Gumenyuk, D.; Dejardin, P.; Yu, S.; Jan, P.; Kubicek, J.; Serrano, C.; Raimond, E.; Dirksen, G.; Ivanov, I.; Groudev, P.; Kowal, K.; Prosek, Andrej; Nitoi, M.; Vitazkova, J.; Hirata, K.; Burgazzi, L.

    2016-01-01

    impact, size of rooms on the path etc. In any case the impact of very hot gas and of hydrogen has to be considered. The dependencies between reactor accident and SFP management appear to be an important issue for L2 PSA risk assessment. The report provides information on ongoing R and D activities which may support the preparation of guidelines for 'traditional' and extended L2 PSA. In addition, a list is provided for those topics which seem to have inadequate covering in present activities. Appendices cover the level 1 shutdown states PSA and country-specific examples related to shutdown PSA and spent fuel pool PSA from ASAMPSA-E WP40 contributing organizations. (authors)

  8. FBR type reactor

    International Nuclear Information System (INIS)

    Jinbo, Masakazu; Kawakami, Hiroto; Nagaoka, Kazuhito.

    1996-01-01

    In a LMFBR type reactor, a liquid level control means is disposed for lowering a level of liquid metal present in an annular gap along with temperature elevation of the liquid metal after the level is once elevated upon start-up of the reactor. In addition, a liquid level measuring means is disposed for measuring the level of the liquid metal present in the annular gap so as to intermittently lower the liquid level. Thus, temperature gradient in the vertical direction of the container can be moderated compared with the case where the liquid level is not changed or the case where temperature is changed together with the elevation of the liquid level. As a result, the change of difference of thermal expansion is decreased to reduce stresses generated in the circumferential direction thereby preventing occurrence of a liquid level heat ratchet phenomenon. Even if the liquid level control means should stop during operation, the liquid level lowers and does not cause a sharp heat gradient as in the case where the liquid level is elevated, and since the temperature of the liquid level is lowered even after shut down of the reactor, generated stresses are not increased. Safety of an intermediate heat exchanger vessel is ensured and observation from a control chamber is enabled. (N.H.)

  9. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  10. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  12. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  13. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  14. Calibration of RB reactor power

    International Nuclear Information System (INIS)

    Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

    1976-09-01

    The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

  15. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  17. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  18. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  19. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  1. A high-speed data acquisition system to measure low-level current from self-powered flux detectors in CANDU nuclear reactors

    International Nuclear Information System (INIS)

    Lawrence, C.B.; Hall, D.S.

    1982-05-01

    Self-powered flux detectors are used in CANDU nuclear power reactors to determine the spatial neutron flux distribution in the reactor core for use by both the reactor control and safety systems. To establish the dynamic response of different types of flux detectors, the Chalk River Nuclear Laboratories have an ongoing experimental irradiation program in the NRU research reactor for which a data acquistion system has been developed. The system described in this paper is used to measure the currents from the detectors both at a slow, regular logging interval, and at a rapid, adaptive rate following a reactor shutdown. Currents that range from 100 pA to 1 mA full scale can be measured from up to 38 detectors and stored at sampling rates of up to 20 samples per second. The dynamic characteristics of the detectors can be computed from the stored records. The data acquisition system comprises a DEC LSI-11/23 microcomputer, dual cartridge disks, floppy disks, a hard copy and a video display terminal. The RT-11 operating system is used and all application programs are written in FORTRAN

  2. Reactor control device

    International Nuclear Information System (INIS)

    Kameda, Akiyuki.

    1979-01-01

    Purpose: To enable three types of controls, that is, level control, scram control and excess reactivity control required for a reactor by a same mechanism by feeding neutron absorber liquid and pressure control gas to several blind pipes provided in the reactor core. Constitution: A plurality of blind pipes are disposed spaced apart in a reactor core and connected by way of injection pipes to a neutron absorber liquid tank. A pressure regulator is connected to the blind pipes, to which pressure control gas is supplied. The neutron absorber liquid used herein consists of sodium, potassium or their alloy, or mercury as a basic substance incorporated with one or more selected from boron, tantalum, rhenium, europium or their compounds. The level control, scram control and excess reactivity control can be attained by moderating the pressure changes in the pressure control gas or by regulating the fluctuation in the liquid level. (Horiughi, T.)

  3. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  4. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  5. Geochemical behaviour study of radionuclides and their radiogenic daughters in the vicinity of Oklo 10 and 13 natural nuclear reactors (Gabon) - Application to high-level radioactive waste disposal

    International Nuclear Information System (INIS)

    Menet-Dressayre, Catherine

    1992-01-01

    Since 1981, the discovery of new and almost unaltered natural nuclear reactors in the uranium mine of Oklo (Gabon) renewed the interest of scientific community. Indeed, due to their specific features, these reactors could be extensively investigated as natural analogues to better understand the geochemical processes which may occur in a high level nuclear waste repository. The aim of this PhD thesis is to determine the present distribution of a few radionuclides or their radiogenic daughters initially formed within the reaction zones and to infer their geochemical behaviour, subsequently to the stopping of nuclear reactions. Our study was focused on reactors 10 and 13 and their immediate sandstone surroundings in order to decipher the fate of U, Y and light rare earth elements which are assumed to be chemical analogues of actinides and fission products. Mineralogical observations, chemical and isotopic analyses on bulk rocks, led us to conclude that a part of radionuclides, as well as their daughters, remained confined within the reactions zones, in association with secondary mineral phases, whereas another part migrated towards tbe reactor rims. The radionuclides were concentrated at the reactor border or migrated within the first few metres of the surrounding sandstone, according to the intensity of nuclear reactors and the presence of the so-called 'facies argile de pile' which constitutes an intermediate facies between that of reactor cores and that of the surrounding sandstone. In the latter, long range elemental transfers occurred via fissures. Some of them, contemporaneous to the nuclear reactions drained radionuclides-rich fluids at temperatures of about 150-170 deg. C. More recent fissures, observed only in the environment of reactor 13, have allowed the transport of hotter hydrothermal fluids (about 310 deg. C), likely related to the nearby intrusion of dolerite dyke. The principal implications of this work for the disposal of nuclear wastes

  6. Removal of Legacy Low-Level Waste Reactor Moderator De-ionizer Resins Highly Contaminated with Carbon-14 from the 'Waste with no Path to Disposal List' Through Innovative Technical Analysis and Performance Assessment Techniques

    International Nuclear Information System (INIS)

    Goldston, W.T.; Hiergesell, R.A.; Kaplan, D.I.; Pope, H.L.

    2006-01-01

    At the Savannah River Site (SRS), nuclear production reactors used de-ionizers to control the chemistry of the reactor moderator during their operation to produce nuclear materials primarily for the weapons program. These de-ionizers were removed from the reactors and stored as a legacy waste and due to the relatively high carbon-14 (C-14) contamination (i.e., on the order of 740 giga becquerel (GBq) (20 curies) per de-ionizer) were considered a legacy 'waste with no path to disposal'. Considerable progress has been made in consideration of a disposal path for the legacy reactor de-ionizers. Presently, 48 - 50 de-ionizers being stored at SRS have 'no path to disposal' because the disposal limit for C-14 in the SRS's low-level waste disposal facility's Intermediate Level Vault (ILV) is only 160 GBq (4.2 curies) per vault. The current C-14 ILV disposal limit is based on a very conservative analysis of the air pathway. The paper will describe the alternatives that were investigated that resulted in the selection of a route to pursue. This paper will then describe SRS's efforts to reduce the conservatism in the analysis, which resulted in a significantly larger C-14 disposal limit. The work consisted of refining the gas-phase analysis to simulate the migration of C-14 from the waste to the ground surface and evaluated the efficacy of carbonate chemistry in cementitious environment of the ILV for suppressing the volatilization of C-14. During the past year, a Special Analysis was prepared for Department of Energy approval to incorporate the results of these activities that increased the C-14 disposal limits for the ILV, thus allowing for disposal of the Reactor Moderator De-ionizers. Once the Special Analysis is approved by DOE, the actual disposal would be dependent on priority and funding, but the de-ionizers will be removed from the 'waste with no path to disposal list'. (authors)

  7. Influence of substrates on nitrogen removal performance and microbiology of anaerobic ammonium oxidation by operating two UASB reactors fed with different substrate levels

    International Nuclear Information System (INIS)

    Tang Chongjian; Zheng Ping; Hu Baolan; Chen Jianwei; Wang Caihua

    2010-01-01

    Both ammonium and nitrite act as substrates as well as potential inhibitors of anoxic ammonium-oxidizing (Anammox) bacteria. To satisfy demand of substrates for Anammox bacteria and to prevent substrate inhibition simultaneously; two strategies, namely high or low substrate concentration, were carefully compared in the operation of two Anammox upflow anaerobic sludge blanket (UASB) reactors fed with different substrate concentrations. The reactor working at relatively low influent substrate concentration (NO 2 - N, 240 mg-N L -1 ) was shown to avoid the inhibition caused by nitrite and free ammonia. Using the strategy of low substrate concentration, a record super high volumetric nitrogen removal rate of 45.24 kg-N m -3 day -1 was noted after the operation of 230 days. To our knowledge, such a high value has not been reported previously. The evidence from transmission electron microscopy (TEM) showed that the morphology and ultrastructure of the Anammox cells in both the reactor enrichments was different.

  8. The aqueous homogeneous suspension reactor project

    International Nuclear Information System (INIS)

    1975-01-01

    The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described

  9. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  10. Small propulsion reactor design based on particle bed reactor concept

    International Nuclear Information System (INIS)

    Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

    1989-01-01

    In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

  11. RB Research nuclear reactor, Annual report for 2004

    International Nuclear Information System (INIS)

    Dasic, N.; Pesic, M.; Nikolic, D; Jevremovic, M.; Eskirovic, B.

    2005-02-01

    Report on RB reactor operation during 2004 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. It contains data about reactor operation during previous 8 years. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation, Annex 1. contains data about heavy water degradation, and Annex 2 is the certificate about the crane bridge in the reactor hall

  12. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  13. Molten salt reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  14. LWR type reactor

    International Nuclear Information System (INIS)

    Kato, Kiyoshi.

    1993-01-01

    A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  16. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  17. Reactor utilization; Eksploatacija reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel.

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  19. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  20. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  1. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  2. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  4. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  5. Hydrodynamics of multi-phase packed bed micro-reactors

    NARCIS (Netherlands)

    Márquez Luzardo, N.M.

    2010-01-01

    Why to use packed bed micro-reactors for catalyst testing? Miniaturized packed bed reactors have a large surface-to-volume ratio at the reactor and particle level that favors the heat- and mass-transfer processes at all scales (intra-particle, inter-phase and inter-particle or reactor level). If the

  6. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  7. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  8. Dosimetry system of the RB reactor

    International Nuclear Information System (INIS)

    Lolic, B.; Vukadin, D.

    1962-01-01

    Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor

  9. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  10. Building Newcomer Competence for NPP Safety Assessment through Learning by Doing: Development of Level 1 Probabilistic Safety Assessment for Research Reactors

    International Nuclear Information System (INIS)

    Kuzmina, Irina

    2014-01-01

    Final remarks: • COMPASS-M project is a very fruitful study. 1. State-of-the-art competence for PSA technique in Malaysia (applicable to nuclear installations, incl. RR and NPP). 2. PSA model and report for the operating research reactor in Malaysia. → Risk estimate of core damage and ranking contributors to the risk; → Basis for further safety improvement of RR as appropriate. 3. Input for IAEA’s publications on PSA for research reactors. • The results will be available to interested Member States (security considerations be addressed); → Completion in mid-2014, paper to be published in PSAM-12; ► Managerial support is instrumental for success of learning-by-doing projects

  11. Variation of sup 137 Cs levels between sexes, body sizes and collection localities of mosquitofish, Gambusia holbrooki (Girard 1859), inhabiting a reactor cooling reservoir

    Energy Technology Data Exchange (ETDEWEB)

    Newman, M.C.; Brisbin, I.L. Jr. (Savannah River Ecology Lab., Aiken, SC (USA))

    1990-01-01

    Whole body concentrations of {sup 137}Cs were determined for 190 mosquitofish (Gambusia holbrooki) collected in April 1987 from three locations in a former cooling reservoir which had been contaminated with {sup 137}Cs from production reactor effluents between 1961 and 1964. Male fish collected near the point where the reactor effluent had entered the reservoir tended to have higher {sup 137}Cs concentrations than those from the other locations. Females did not differ in {sup 137}Cs concentrations between the three locations. Females at the site of contaminant entry tended to decrease in {sup 137}Cs concentration as body size increased. These results suggest that that radionuclide whole body concentrations may vary in unexpected ways between sex or size classes within a given species and that such differences may also vary within microgeographic scales. (author).

  12. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  13. Influence of substrates on nitrogen removal performance and microbiology of anaerobic ammonium oxidation by operating two UASB reactors fed with different substrate levels

    Energy Technology Data Exchange (ETDEWEB)

    Tang Chongjian [Department of Environmental Engineering, Zhejiang University, Hangzhou 310029 (China); Zheng Ping, E-mail: pzheng@zju.edu.cn [Department of Environmental Engineering, Zhejiang University, Hangzhou 310029 (China); Hu Baolan; Chen Jianwei; Wang Caihua [Department of Environmental Engineering, Zhejiang University, Hangzhou 310029 (China)

    2010-09-15

    Both ammonium and nitrite act as substrates as well as potential inhibitors of anoxic ammonium-oxidizing (Anammox) bacteria. To satisfy demand of substrates for Anammox bacteria and to prevent substrate inhibition simultaneously; two strategies, namely high or low substrate concentration, were carefully compared in the operation of two Anammox upflow anaerobic sludge blanket (UASB) reactors fed with different substrate concentrations. The reactor working at relatively low influent substrate concentration (NO{sub 2}{sup -}N, 240 mg-N L{sup -1}) was shown to avoid the inhibition caused by nitrite and free ammonia. Using the strategy of low substrate concentration, a record super high volumetric nitrogen removal rate of 45.24 kg-N m{sup -3} day{sup -1} was noted after the operation of 230 days. To our knowledge, such a high value has not been reported previously. The evidence from transmission electron microscopy (TEM) showed that the morphology and ultrastructure of the Anammox cells in both the reactor enrichments was different.

  14. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    Banerjee, S.; Kamath, H.S.

    2005-01-01

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  15. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

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

  17. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  18. Reactor container facility

    International Nuclear Information System (INIS)

    Saito, Takashi; Nagasaka, Hideo.

    1990-01-01

    A dry-well pool for spontaneously circulating stored pool water and a suppression pool for flooding a pressure vessel by feeding water, when required, to a flooding gap by means of spontaneous falling upto the flooding position, thereby flooding the pressure vessel are contained at the inside of a reactor container. Thus, when loss of coolant accidents such as caused by main pipe rupture accidents should happen, pool water in the suppression pool is supplied to the flooding gap by spontaneously falling. Further, if the flooding water uprises exceeding a predetermined level, the flooding gap is in communication with the dry-well pool at the upper and the lower portions respectively. Accordingly, flooding water at high temperature heated by the after-heat of the reactor core is returned again into the flooding gap to cool the reactor core repeatedly. (T.M.)

  19. Reactor fuel charging equipment

    International Nuclear Information System (INIS)

    Wade, Elman.

    1977-01-01

    In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

  20. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1972-06-01

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs