Annual report of Department of Research Reactor and Tandem Accelerator, JFY2013. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Kashima, Yoichi; Murayama, Yoji; Nakamura, Kiyoshi; Uno, Yuki; Hirane, Nobuhiko; Ohuchi, Hitoshi; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Harada, Yuko; Kadokura, Masakazu; Machi, Sumire; Takita, Chiharu

2015-02-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2013. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

Safety design concept and analysis for the upgrading JRR-3

International Nuclear Information System (INIS)

Onishi, N.; Isshiki, M.; Takahashi, H.; Takayanagi, M.

1990-01-01

The Research Reactor No.3 (JRR-3) is under reconstruction for upgrading. This paper describes the safety design concepts of the architectural and engineering design, anticipated operational transients and accident conditions which are the postulated initiating events for the safety evaluation, and the safety criteria of the upgraded JRR-3. The safety criteria are defined taking into account those of Light Water Reactors and the characteristics of the research reactor. Using the example of the safety analysis, this paper describes analytical results of a reactivity insertion by removal of in-core irradiation samples, a pipeline break at the primary coolant loop and flow blockage to a coolant channel, which are the severest postulated initiating events of the JRR-3

Overhaul of the heat exchanger in JRR-3

International Nuclear Information System (INIS)

Ouchi, Yasuhiro; Kawamata, Satoshi; Taguchi, Yuji; Kamiishi, Eigo; Koda, Nobuyuki

2013-01-01

In JRR-3, heat exchangers are installed in the cooling system equipment to remove the heat generated in the nuclear reactor, For the heat exchangers, overhaul inspection based on the JRR-3 reactor facility maintenance plan, as well as the inspection and maintenance based on reactor facility security provisions and JRR-3 operation guidelines are systematically conducted. Considering the results of overhaul inspection, the second overhaul inspection was applied to the primary coolant heat exchanger. The thinning of heat transfer tubes is within judgment standards with little effects of aging, which verified their soundness. From the fact that the effects of corrosion have been confirmed on the inside of the water chamber, repair work through overlay welding or the like is planned in the next overhaul. As for heavy water heat exchanger and the spent fuel pool water heat exchanger, it is planned to conduct the second overhaul inspection in FY2013 to confirm their soundness. (A.O.)

Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Energy Technology Data Exchange (ETDEWEB)

Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

1985-07-01

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2012. Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Murayama, Yoji; Ishii, Tetsuro; Nakamura, Kiyoshi; Uno, Yuki; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Odauchi, Shouji; Maruo, Takeshi

2014-03-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2012 and March 31, 2013. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for department of research reactor and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, outcomes in service and technical developments and so on. (author)

Estimation of reactor pool water temperature after shutdown in JRR-3M

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

Current status of JRR-3. After the 3.11 earthquake

International Nuclear Information System (INIS)

Arai, Masaji; Murayama, Yoji; Wada, Shigeru

2012-01-01

JRR-3 at Tokai site of JAEA was in its regular maintenance period, when the Great East Japan Earthquake was taken place on 11th March 2011. The reactor building with their solid foundations and the equipment important to safety survived the earthquake without serious damage and no radioactive leakage has been occurred. Recovery work is planned to be completed by the end of this March. At the same time, check and test of the integrity of all components and seismic assessment to show resistance with the 3.11 earthquake have been carried out. JRR-3 will restart its operation after completing above mentioned procedures. (author)

Long-period maintenance program of JRR-3M

International Nuclear Information System (INIS)

Murayama, Yoji; Onozaki, Michio; Kakefuda, Kazuhiro

1999-01-01

The JRR-3M is a swimming pool type research reactor with maximum thermal output of 20 MW, classed in a large-scale research reactor and composed of numerous facilities. An operation cycle consists of 4 weeks of continuously reactor operation and 1 week of shutdown work such as fuel exchange and irradiation sample handling. The temporary and small-scale maintenance is also carried out during the shutdown period. Systematic maintenance period through a year usually follows 2 stages. The former stage of about 4 weeks is prepared to maintain the general facilities related to the overall reactor system such as computer system. The latter stage of about 9 weeks is prepared to maintain the other individual facilities such as cooling facilities and control facilities that are to be inspected by the regulatory body. Since preventive maintenance is required in order to achieve an improvement of safety, reliability and availability throughout the life cycle of the reactor, a long-period maintenance program should be prepared properly in consideration of an ageing degradation of the facilities. The long-period maintenance program of JRR-3M, which determines not only the replacement frequency but also the maintenance frequency of the facilities, is made on the basis of an ageing degradation of the facilities, irradiation effects of the materials, recommendations from the manufacturer and JRR-3M maintenance records. The replacement frequency recommended by the manufacturer is examined from the viewpoint of the safety significance of the facilities and availability of the components. The program is arranged so that it may not concentrate the replacement of the facilities in the specific fiscal year and smooth the budget and maintenance period throughout the life cycle of the reactor. There have been two unanticipated failures of the facilities, which required rearranging the program. (author)

JRR-3 maintenance program utilizing accumulated maintenance data

International Nuclear Information System (INIS)

Izumo, Hironobu; Kato, Tomoaki; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

2007-07-01

JRR-3(Japan Research Reactor No.3) has been operated for more than 15 years after the modification, without significant troubles by carrying out maintenance such as the preventive maintenance (mainly time-based maintenance) for the safety-grade equipments and the breakdown maintenance for the non-safety-grade equipments. Unscheduled shutdowns causes by aged non-safety-grade equipments have been increasing, and the resources such as budgets have been decreasing year by year. In this situation, JRR-3 maintenance program was reviewed about safety, reliability and economic efficiency. This report offers the policy of the maintenance review and the future direction of maintenance programs. (author)

Reactivity initiated accident analyses for the safety assessment of upgraded JRR-3

International Nuclear Information System (INIS)

Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

1984-08-01

JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. This report describes the analytical results of reactivity initiated accidents for the safety assessment of upgraded JRR-3. The following five cases have been selected for the assessment; (1) uncontrolled control rod withdrawal from zero power, (2) uncontrolled control rod withdrawal from full power, (3) removal of irradiation samples, (4) increase of primary coolant flow, (5) failure of heavy water tank. Parameter studies have been made for each of the above cases to cover possible uncertainties. All analyses have been made by a computer code EUREKA-2. The results show that the safety criteria for upgraded JRR-3 are all met and the adequacy of the design is confirmed. (author)

Inspection and repair in JRR-3 after the 2011 off the Pacific coast of Tohoku Earthquake

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Hosoya, Toshiaki; Nagadomi, Hideki; Torii, Yoshiya

2014-01-01

In the 2011 off the Pacific Coast of Tohoku Earthquake, seismic intensity of 6 lower was observed at Tokai Village. However, the maximum acceleration of ground motion that was observed in the JRR-3 reactor facilities had exceeded the maximum response acceleration at the time of design. Therefore, to confirm whether the predetermined performance of the facility equipment of the reactor facilities had been maintained after the earthquake, soundness confirmation inspection was carried out. In the inspection, the soundness of equipment and facilities was evaluated from the results of the equipment inspection and seismic impact assessment, and the repair work was applied when necessary. As a result, it was confirmed that after the earthquake, the equipment of JRR-3 reactor facilities maintained the predetermined performance, and was possible to resume operation. The following item are reported here: (1) overview of JRR-3, (2) conditions of JRR-3 reactor facilities while earthquake occurrence, (3) basic principle for soundness evaluation of facilities, (4) soundness confirmation of buildings and structures, (5) contents of repair, and (6) soundness verification and comprehensive evaluation of each facility and equipment. (A.O.)

Experiences on operation, maintenance and utilization in JRR-2

International Nuclear Information System (INIS)

1994-08-01

The Japan Research Reactor No.2 (JRR-2) is a high performance 10 MW multi purpose research reactor, heavy water moderated and cooled enriched uranium fuel used. Since the first criticality was attained in October, 1960, JRR-2 has been operated to satisfy the utilization demands, such as irradiation of fuel and materials, neutron beam experiments, radio isotope production and B.N.C.T (Boron Neutron Capture Therapy). During the operation, various kinds of troubles mainly caused by the old design concept had been occurred at the JRR-2 systems and components. Those troubles were solved with adequate countermeasures of timely repairs and large scale modifications with newest techniques. The works above were completely carried out by the staff of JRR-2 and related divisions. As a result, JRR-2 became one of the oldest research reactors which are still under operation in the world. Since JRR-2 has been utilized for more than 30 years, the operation mode was changed from 12 days-one cycle to 3 days-one cycle in April, 1994, taking into consideration aging of the reactor systems. In this paper, the experiences of JRR-2 for more than 30 years such as operation, maintenance, repair, modifications and utilization on JRR-2 are described. (author)

Measurements of neutron fluxes and cadmium ratio at equilibrium core in JRR-3M

International Nuclear Information System (INIS)

Ohtomo, Akitoshi; Sasajima, Fumio; Ishida, Takuya; Shigemoto, Masamitsu; Takahashi, Hidetake; Maejima, Takeshi; Sekine, Katsunori.

1993-08-01

Construction and characteristics tests of JRR-3M (Modified JRR-3) had been completed on October 1990, and the reactor reached to equilibrium core in July 1991. Measurements of neutron flux and cadmium ratio in Hydraulic irradiation facility (HR) and Pneumatic irradiation facility (PN) at 20 MW reactor power were carried out for the equilibrium core from May to August 1991 and for the latest core in April 1993. The results at the equilibrium core and the latest core are described in this paper. (author)

Core burn-up calculation method of JRR-3

International Nuclear Information System (INIS)

Kato, Tomoaki; Yamashita, Kiyonobu

2007-01-01

SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

Kakefuda, K.; Tani, M.; Isshiki, M.

1992-01-01

The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

Enhancement of control rod drive mechanism seating position detector for JRR-3

International Nuclear Information System (INIS)

Ohuchi, Satoshi; Kurumada, Osamu; Kamiishi, Eigo; Sato, Masayuki; Ikekame, Yoshinori; Wada, Shigeru

2016-06-01

The purpose of the control rod drive mechanism seating position detector for JRR-3 is one of methods for confirming the shutdown condition of the reactor and sending out the seat position signal to other systems. The detector has been utilizing more than 25 years with maintenance regularly. However, some troubles occurred recently. Moreover, the detector has already been discontinued, and it is confirmed that the successor detector is unsuitable for the control rod drive mechanism of JRR-3. Therefore, it was necessary to select the adequate detector to the control rod drive mechanism of JRR-3. Accordingly, we built a test device with the aim of verifying several detectors for integrity and function. At the time of the test for performance confirmation, it was occurred unexpected problems. Nevertheless, we devise improvement of the problems and took measures. Thus we were able to make adequate detector for JRR-3 and replace to enhanced detector. This paper reports the Enhanced of Control rod drive mechanism seating position detector. (author)

Characteristics of the JRR-3M neutron guide tubes

International Nuclear Information System (INIS)

Suzuki, Masatoshi; Ichikawa, Hiroki; Kawabata, Yuji.

1993-01-01

Large scale neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No.3, JRR-3M). The total length of the guide tubes is 232m. The neutron fluxes and spectra were measured at the end of the neutron guide tubes. The neutron fluxes of thermal neutron guide tubes with characteristic wavelength of 2A are 1.2 x 10 8 n/cm 2 · s. The neutron fluxes of cold guide tubes are 1.4 x 10 8 n/cm 2 · s with characteristic wavelength of 4A and 2.0 x 10 8 n/cm 2 · s with 6A when the cold neutron source is operated. The neutron spectra measured by time-of-flight method agree well with their designed ones. (author)

Development of heat transfer package for core thermal-hydraulic design and analysis of upgraded JRR-3

International Nuclear Information System (INIS)

Sudo, Yukio; Ikawa, Hiromasa; Kaminaga, Masanori

1985-01-01

A heat transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Reactor-3 (JRR-3) which is to be remodeled to a 20 MWt pool-type, light water-cooled reactor with 20 % low enriched uranium (LEU) plate-type fuel. This paper presents the constitution of the developed heat transfer package and the applicability of the heat transfer correlations adopted in it, based on the heat transfer experiments in which thermal-hydraulic features of the new JRR-3 core were properly reflected. (author)

The cold neutron facility of the JRR-3M

International Nuclear Information System (INIS)

Kumai, T.; Suzuki, M.; Kakefuda, K.

1992-01-01

A description is given of a cold neutron source and neutron guide tubes of the JRR-3M. The installation of the cold neutron source (CNS) together with the neutron guide system is one of the principal objectives of the remodeling project of the JRR-3 and this CNS is the first one that was installed in the high neutron flux reactors of 14 orders of magnitude in Japan. The CNS is a liquid hydrogen moderator and vertical thermosyphon type. It mainly consists of a hydrogen plant for liquid hydrogen and helium refrigerator plant for cold helium gas. Five neutron guide tubes are installed to get thermal and cold neutron beams in the beam hall. The CNS and the guide tubes have been operated very well since August 1990. (author)

Decommissioning program of JRR-2

International Nuclear Information System (INIS)

Kishimoto, Katsumi; Banba, Masao; Arigane, Kenji

1999-01-01

Japan Research Reactor No.2(JRR-2), heavy water moderated and cooled tank type research reactor with maximum thermal power of 10 MW, was used over 36 years, and was permanently shut down in December, 1996. Afterward, dismantling report was submitted to the STA, and dismantling was begun in 1997. Decommissioning of JRR-2 is planned in 11 years from 1997 to 2007, and the program is divided into 4 phases. Phase 1 had already been ended, phase 2 is being executed at present. Reactor body will be removed in phase 4 by one piece removal or caisson techniques. On reactor building, it is planned to use effectively as a hot experimental facilities after decommissioning ends. How to treat heavy water and primary cooling system contaminated by tritium becomes an important problem to lead decommissioning to success because JRR-2 is heavy water reactor. On heavy water, transportation to foreign country is planned in phase 2. On primary cooling system, it is planned to remove and dispose the majority in phase 3, and tritium decontamination with technique established by the proof test is planned before them. As a preparation for them, various investigation and examination are being advanced at present. (author)

Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-03-01

JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

Full core operation in JRR-3 with LEU fuels

International Nuclear Information System (INIS)

Murayama, Y.; Issiki, M.

1995-01-01

The new JRR-3 a 20MWT swimming pool type research reactor, is made up of plate type LEU fuel elements with U-Al x fuel at 2.2 gU/cm 3 . Reconstruction work for the new JR-3 was a good success, and common operation started in November 1990, and 7 cycles (26 days operation/cycle) have passed. We have no experience in using such a high uranium density fuel element with aluminide fuel. So we plan to examine the condition of the irradiated fuel elements with three methods, that is, measurement of the value of FFD in operation, observation of external view of the fuels in refueling work and postirradiation examination after maximum burn-up will be established. In the results of the first two methods, the fuel elements of JRR-3 is burned up normally and have no evidence of failure. (author)

Status of operation and utilization of JRR-4

International Nuclear Information System (INIS)

Tsuruta, H.; Niiho, T.; Matsuura, H.; Iida, K.; Ikeda, Y.; Funayama, Y.

1988-01-01

Japan Research Reactor No.4 (JRR-4) is a swimming pool type research reactor, 93% enriched uranium ETR-type fuel used, light water moderated and cooled. It was initially constructed to pursue mainly shielding experiments for nuclear ship. Today, it is widely utilized for many researches and experiments and reactor operation training of Nuclear Engineering School students. The construction of JRR-4 started in June 1962 and the initial criticality reached on January 28, 1965. The maximum power of JRR-4 was raised to 3500 kW in October 1976. The reactor has been operated satisfactorily for about 23 years since the initial criticality. In this paper, a general description is given on the JRR-4 including future plans. (author)

Replacement of the JRR-3 computer system

Energy Technology Data Exchange (ETDEWEB)

Kato, Tomoaki; Kobayashi, Kenichi; Suwa, Masayuki; Mineshima, Hiromi; Sato, Mitsugu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-10-01

The JRR-3 computer system contributes to stable operation of JRR-3 since 1990. But now about 10 years have passed since it was designed and some problems have occurred. Under these situations, we should replace the old computer system for safe and stable operation. In this replacement, the system is improved as regards man-machine interface and efficiency about maintenance. The new system consists of three functions, which are 'the function of management for operation information' (renewal function), 'the function of management for facility information' (new function) and the function of management for information publication' (new function). By this replacement, New JRR-3 computer system can contribute to safe and stable operation. (author)

Replacement of the JRR-3 computer system

International Nuclear Information System (INIS)

Kato, Tomoaki; Kobayashi, Kenichi; Suwa, Masayuki; Mineshima, Hiromi; Sato, Mitsugu

2000-01-01

The JRR-3 computer system contributes to stable operation of JRR-3 since 1990. But now about 10 years have passed since it was designed and some problems have occurred. Under these situations, we should replace the old computer system for safe and stable operation. In this replacement, the system is improved as regards man-machine interface and efficiency about maintenance. The new system consists of three functions, which are 'the function of management for operation information' (renewal function), 'the function of management for facility information' (new function) and the function of management for information publication' (new function). By this replacement, New JRR-3 computer system can contribute to safe and stable operation. (author)

Developments of high-performance moderator vessel for JRR-3 cold neutron source

International Nuclear Information System (INIS)

Arai, Masaji; Tamura, Itaru; Hazawa, Tomoya

2015-05-01

The cold neutron source (CNS) facility converts thermal neutrons into cold neutrons to moderate neutrons with liquid hydrogen. The cold neutron beam at Japan Research Reactor No. 3 (JRR-3) is led to the beam experimental devices in the beam hall through neutron guide tubes. High intensities of the cold neutron beam are always demanded for increasing the experimental effectiveness and accuracy. In the Department of Research Reactor and Tandem Accelerator, developments of high-performance CNS moderator vessel that can produce cold neutron intensity about two times higher compared to the existing vessel have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the design and manufacture of new vessel. In the present study, design strength evaluation, mockup test, simulation for thermo-fluid dynamics of the liquid hydrogen and strength evaluation of the different-material-bonding were studied. By these evaluation results, we verified that the developed new vessel can be applied to CNS moderator vessel of JRR-3. (author)

Evaluation on activation activity of reactor in JRR-2 applied 3 dimensional model to neutron flux calculation

International Nuclear Information System (INIS)

Kishimoto, Katsumi; Arigane, Kenji

2005-03-01

Revaluation to activation activity of reactor evaluated at the notification of dismantling submitted in 1997 was carried out in JRR-2 where decommissioning was advanced now. In the revaluation, estimation accuracy on neutron streaming at various horizontal experimental tubes was improved by applying 3 dimensional model to neutron transport calculation that had been carried out by 2 dimensional model, and calculating with TORT. As the result, excessive overestimations on horizontal experimental tubes and biological shield that had greatly contributed to total activation activity in evaluation at the notification of dismantling was revised, sum of their activation activities in the revaluation decreased to 1/18 (case after 1 year from the permanent shutdown of reactor) of evaluation at the notification of dismantling, and the structural materials that had large activation activity were changed. By the above, it was shown that introducing 3 dimensional model was effective in evaluation on activation activity of the research reactor that had a lot of various experimental tubes. Total activation activity of reactor by the revaluation depended on control rods, thermal shield plates and horizontal experimental tubes, and the value after 1 year from the permanent shutdown of reactor was 1.9x10 14 Bq. (author)

Modification of JRR-4 based on safety evaluation

Energy Technology Data Exchange (ETDEWEB)

Izumo, Hironobu; Nakajima, Teruo; Funayama, Yoshiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

Since the first criticality was achieved on January 28, 1965, JRR-4 has been operated safely until on January 12, 1996. The modification of JRR-4 was planned according to the framework of reduced enrichment on research reactor program. The modification was designed based on the several national safety guides. JRR-4 has some modifications of facilities to satisfy the guides and guides criteria. (author)

Advancement of neutron radiography technique in JRR-3M

International Nuclear Information System (INIS)

Matsubayashi, Masahito

1999-01-01

The JRR-3M thermal neutron radiography facility (JRR-3M TNRF) was completed in the JRR-3M of the Japan Atomic Energy Research Institute in 1991 and has been utilized as research tools for various kinds of research fields such as thermal hydraulic researches, agricultural researches, medical researches, archaeological researches and so on. High performance of the JRR-3M TNRF such as high neutron flux, high collimator ratio and wide radiographing field has enabled advanced researches and stimulated developments of advanced neutron radiography (NR) systems for higher spatial resolution and for higher temporal resolution. Static NR systems using neutron imaging plates or cooled CCD camera with high spatial resolution, a real-time NR system using a silicon intensifier target tube camera and a high-frame-rate NR system using a combination of an image intensifier and a high speed digital video camera with high temporal resolution have been developed to fill the requirements from researchers. (author)

Comparisons of fission track age-determination and radioactivation analysis among radiation fields. Characteristic values for pneumatic tube, JRR-3M PN2

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Tatsuro [Kagoshima Univ. (Japan). Faculty of Education; Tomura, Kenji

1996-01-01

It is necessary for Fission track age determination and radioactivation analysis to define the energy spectrum of a reactor neutron as the basis for calculating the nuclear reaction rate. This study was made aiming to define the spectrum for PN2 pneumatic tube of No. 3 plant of Japan Atomic Energy Research Institute (JAERI). Characteristics such as conventional flux, cross reaction, neutron temperature and integral fast flux were examined for 3 facilities; JRR-2, JRR-4 and JRR-3M. And the neutron spectrum for each facility was estimated based on the assumptions; Maxwell distribution for thermal neutrons, 1/E for intermediate neutrons and Watt`s formula for fast neutrons. The present results showed that the spectrum for JRR-3M, PN2 was discontinuous in the range from the intermediate to the fast neutron, whereas those for other facilities were found considerably improved. (M.N.)

The reduced enrichment program for JRR-4

International Nuclear Information System (INIS)

Takayanagi, M.

1992-01-01

Japan Research Reactor No. 4(JRR-4) with the rated power of 3.5 MW, swimming pool type research reactor, 93 % enriched uranium ETR-type fuel used, light water moderated and cooled. The first criticality reached on 28th January, 1965. The reactor has operated for about 26 years. However, it was planed to the reduced enrichment of the fuels to low enrichment according to the International Reduced Enrichment for Research and Test Reactors (RERTR) program. This paper describes the program for conversion of the enrichment of fuel from 93 % to less than 20 %. (author)

Status of core conversion with LEU silicide fuel in JRR-4

Energy Technology Data Exchange (ETDEWEB)

Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

1997-08-01

Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10{sup 13}(n/cm{sup 2}/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities.

Status of core conversion with LEU silicide fuel in JRR-4

International Nuclear Information System (INIS)

Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji

1997-01-01

Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10 13 (n/cm 2 /s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities

Decommissioning plan and current status of JRR-2

International Nuclear Information System (INIS)

Kishimoto, Katsumi; Banba, Masao; Arigane, Kenji

1999-01-01

Japan Research Reactor No.2 (JRR-2), heavy water moderated and cooled tank type research reactor with maximum thermal power of 10 MW, was used over 36 years, and was permanently shut down in December, 1996. Afterward, dismantling report was submitted to the STA, and dismantling was begun in 1997. Decommissioning of JRR-2 is planned in 11 years, and the program are divided into 4 phases. Phase l had already been ended, phase 2 is being executed at present. Reactor body will be removed in phase 4 by one piece removal or caisson techniques. On reactor building, it is planned to use effectively as a hot experimental facilities after decommissioning ends. For ensuring safety under decommissioning, detailed examination on work method, exposure and radioactive waste quantities is executed on each dismantling in advance. On exposure of worker in phase 1, it was achieved to control lower than the estimate. How to treat tritium contamination also becomes an important problem to achieve ensuring safety. On heavy water, transportation to foreign country is planned in phase 2. On primary cooling system and reactor building concrete, various investigation and examination is being advanced aiming at phase 3 and 4. (author)

Conceptual study of 99Mo production in JRR-3

International Nuclear Information System (INIS)

Hirose, Akira; Komeda, Masao; Kinase, Masami; Sorita, Takami; Wada, Shigeru

2010-06-01

We investigated the production process of 99 Mo, which is parent nuclide of 99m Tc, in JRR-3. 99m Tc is most commonly used as a radiopharmaceutical in the field of nuclear medicine. Currently the supplying of 99 Mo is only dependent on imports from foreign countries, so JAEA is aiming at domestic production of a part of 99 Mo in cooperation with the industrial arena. This report presents the technical study for the production process of 99 Mo by using the neutron radiation method of (n, γ) reaction in JRR-3. (author)

Modification of JRR-2

International Nuclear Information System (INIS)

Miyasaka, Yasuhiko

1978-01-01

This report gives an outline of some of the main modifications carried out around the Reactor Core on the Research Reactor JRR-2, at the Tokai Research Establishment of JAERI. The JRR-2 was shut down in December 1973, to improve it in heavy water leakage from the metal packing between core tank and support ring, corrosion of the lower shielding plug, and fault in the control-rod mechanism. Main modifications were a standing seal weld at the support ring to stop heavy water leakage, replacement of the reactor top shield and improvement of the helium system. The control-rod assemblies and the refueling devices were replaced by the newly designed ones also. In addition to the modification plan, the irradiated air exhaust system was improved to reduce radioactive argon gas release through the stack. Works were completed successfully in September 1975. But a light water leakage occurred at the stand pipe below the light water tank on November 11, 1975, which was repaired in about 4 months. When considering the operation of above 5,000 hours after the modification, however, the quality of the modification work may be said to be quite satisfactory. The present report in which works to the completion are described may be valuable as a record of reactor modification which is a new experience at JAERI. (auth.)

Detailed design of neutron guide tubes at the upgraded JRR-3, (1)

International Nuclear Information System (INIS)

Harami, Taikan; Umemura, Mutsumi; Ebisawa, Tohru.

1985-07-01

JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. Two guide tubes for thermal neutron and three for cold will be installed in the reactor to transport thermal and cold neutrons from the reactor hall to the experiment hall. This describes the neutron guide tube transmission analysis program, NEUGT, which was developed to assess the design of the neutron guide tubes. The input data plotting program, PLOPINE and the output data plotting program, NEUPLOT are presented in the appendix. The NEUGT program not only calculates a neutron transmission and neutron spectra, assuming the Maxwellian spectra at the entrance of a guide tube, but also analyses the effect of abutment errors. This reports the description and the input data manual of the program in the text. Examples of analysis are given in the appendixes. The program is written in the FORTRAN 77 language for FACOM 380. (author)

Core heat transfer experiment for JRR-3 to be upgraded at 20 MWt, 2

International Nuclear Information System (INIS)

Sudo, Yukio; Miyata, Keiichi; Ikawa, Hiromasa; Ohgawara, Masami; Kaminaga, Masanori

1985-09-01

Experiments were carried out to investigate the condition of onset of nucleate boiling (ONB) and the departure from nucleate boiling (DNB) heat flux under forced convection in a vertical rectangular channel, both of which take important roles in the core thermal-hydraulic design of the upgraded JRR-3. This report presents the validity and applicability of the correlations proposed for ONB condition and DNB heat flux, based on the analysis of the experimental results. The upgraded JRR-3 is a low-pressure, low-temperature research reactor and the core heat generation is removed by two cooling modes, one is natural circulation under upflow up to 200 kW and the other is forced circulation under downflow up to 20 MW. Therefore, the difference in heat transfer characteristics between upflow and downflow were investigated in the experiments, which were carried out by using a heated channel properly simulating a subchannel of fuel element because the heat transfer characteristics are considered to be strongly dependent on the configuration of flow channel. (author)

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)

Fuel conversion of JRR-4 from HEU to LEU

International Nuclear Information System (INIS)

Ichikawa, Hiroki; Nakajima, Teruo

1997-01-01

Japanese JRR-4 (Japan Research Reactor No.4) is a pool type, light water moderated and cooled, ETR type fuel reactor used for Shielding experiments, isotope production, neutron activation analyses, Si doping, reactor students training. It acieved first criticality on January 28, 1965 with maximum thermal power 3.5MW. The standard core consistes of 20 Fuel elements, 7 control rods 5 Irradiation holes, neutron source, graphite reflectors. Available thermal flux is 7x1013 n/cm2/s. Within the RERTR program plans are made for core conversion from HEU to LEU

Improvement of JRR-4 core management code system

International Nuclear Information System (INIS)

Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

2000-01-01

In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system

International Nuclear Information System (INIS)

Arigane, Kenji

1987-04-01

The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)

The 33 years of research reactors in JAERI

International Nuclear Information System (INIS)

1990-11-01

The development and utilization of atomic energy in Japan began with the installation of JRR-1 reactor which attained the criticality in August, 1957, thus the third fire was lighted for the first time in Japan. JRR-2 was constructed as a full scale versatile research reactor, which attained the criticality in October, 1960, and since 1962, it has accomplished the role of the reactor for joint utilization. JRR-3 is the first reactor made in Japan by concentrating Japanese technologies in it, to develop and improve Japanese atomic energy technology. It attained the criticality in September, 1962, and has been used as a versatile research reactor. In 1960, Research Reactor Management Department was founded. JRR-4 was constructed as the research reactor for shielding for developing a nuclear-powered ship, which attained the criticality in January, 1965. The first hot laboratory in Japan for carrying out the post-irradiation test on the fuel and materials irradiated in these research reactors was installed in 1961. The JRR-1 was stopped in September, 1968, and is used as the commemorative exhibition hall. The JRR-3 was reconstructed, and attained the criticality in March, 1990, using 20 % enriched uranium fuel. The course of the research reactors for 33 years is reported. (K.I.)

Full Core Burn-up Calculation at JRR-3 with MVP-BURN

International Nuclear Information System (INIS)

Komeda, Masao; Yamamoto, Kazuyoshi; Kusunoki, Tsuyoshi

2008-01-01

Research reactors use a burnable poison to suppress an excess reactivity in the beginning of reactor lifetime. The JRR-3 (Japan Research Reactor No.3) has used cadmium wires of radius 0.02 cm as a burnable poison. This report describes burn-up calculations of plate fuel models and full core models with MVP-BURN, which is a burn-up calculation code using Monte Carlo method and has been developed in JAEA (Japan Atomic Energy Agency). As the results of calculations of plate models, between a model composed of one burn-up region along the radius direction and a model composed of a few burn-up regions along the radius direction, the effective absorption cross section of 113 Cd has had different tendency on reaching approximate 40. day (10000 MWd/t). And as results of calculations of full core model, it has been indicated that k eff is almost same till approximate 80. day (22000 MWd/t) between a model composed of one burn-up region along the vertical direction and a model composed of a few burn-up regions along the vertical direction. However difference of 113 Cd burn-up becomes pronounced and each k eff makes a difference after 80. day. (authors)

Development of JRR-4 operation support system

International Nuclear Information System (INIS)

Takahashi, Hiroki; Yamamoto, Kazuyoshi; Yoritsune, Tsutomu; Arai, Nobuyoshi; Birumachi, Takashi

1999-03-01

The Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, graphite reflected, pool-type reactor with the thermal output of 3500 kW. In February 1996, a work on core modification from high-enriched fuels to low-enriched ones was started and the work on modification was completed in May 1998. As a part of the work, Operation Support System (OSS) was adopted for the purpose of reducing load of operator, preventing operator's error and supplying operation data to users. In this paper, JRR-4 Operation Support System's configuration, characteristics and development process are described. OSS consists mainly of a workstation (WS), μXL and Personal Computer (PC). WS is used for collection of operation data. μXL is used for cooling system control and PC is used for operators' support. All of the program applications used in WS, μXL and PC are designed to make sure of high flexibility for the demands on modification from operators and new operation modes. The result of adopting this system, the consolidation monitoring, automation of part of reactor operation and data management were realized. In addition, the cooling control system was compacted. (author)

Water chemistry management of research reactor in JAERI

Energy Technology Data Exchange (ETDEWEB)

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

1998-10-01

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

Annual report of department of research reactor, 1992

International Nuclear Information System (INIS)

1993-12-01

The department of research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1992 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as well as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

Neutron scattering research at JAERI reactors - past, present and future -

International Nuclear Information System (INIS)

Funahashi, Satoru; Morii, Yukio; Minakawa, Nobuaki

1992-01-01

It was in 1961 that the first neutron scattering experiment was performed in Japan at JRR-2. The start of JRR-3 in 1964 accelerated the neutron scattering activities in Japan. The research in this field in Japan grew up by using these two research reactors. Among them JRR-2 has played an important role because its neutron flux was about seven times higher than that of the old JRR-3. The completion of the new JRR-3M in 1990 made an epoch to the neutron scattering activities in Japan. The long-waited JRR-3M came up to the expectations of the scientists of Japan. It is a realization of the ideal reactor with tangential beam holes, cold source and neutron guides in a large guide hall. The flux at the neutron scattering instruments is about five times higher than that of JRR-2. Utilization of JRR-3M has just started. Twelve neutron scattering machines are running there. The number will increase up to close twenty in a couple of years. (author)

Assessment of irradiation effects on beryllium reflector and heavy water tank of JRR-3M

Energy Technology Data Exchange (ETDEWEB)

Murayama, Yoji; Kakehuda, Kazuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

The JRR-3M, a swimming pool type research reactor with beryllium and heavy water reflectors, has been operated since 1990. Since the beryllium reflectors are close to fuel and receive high fast neutron fluence in a relatively short time, they may be subject to change their dimensions by swelling due mostly to entrapped helium gaseous. This may bend the reflectors to the outside and narrow gaps between the reflectors and the fuel elements. The gaps have been measured with an ultrasonic thickness gage in an annual inspection. The results in 1996 show that the maximum of expansion in the diametral directions was 0.6 mm against 1.6 mm of a managed value for replacement of the reflector. A heavy water tank of the JRR-3M is made of aluminum alloy A5052. Surveillance tests of the alloy have been conducted to evaluate irradiation effects of the heavy water tank. Five sets of specimens of the alloy have been irradiated in the beryllium reflectors where fast neutron flux is higher than that in the heavy water tank. In 1994, one set of specimens had been unloaded and carried out the post-irradiation tests. The results show that the heavy water tank preserved satisfactory mechanical properties. (author)

A study of the modifications of nuclear instrumentation systems for JRR-2

International Nuclear Information System (INIS)

Azim, Mohammad; Horiki, Ooichiro; Sato, Mitsugu

1978-04-01

In this report a comparative study has been carried out between the original A.M.F. design and the modified design for the nuclear instrumentation systems of the Research Reactor JRR-2, at the Tokai Research Establishment of JAERI. Due to a fire accident in the control room, in July 1968, the originally designed nuclear instrumentation systems, using conventional vacuum tube circuits, were destroyed and were replaced by the modified design, incorporating solid state linear integrated circuits as basic circuit components. The results of the reactor instrumentation systems modification at JRR-2 are very encouraging as the operating efficiency of the Reactor registered an improvement of 43%. Moreover the safety aspects have been fully taken care of in the new design and the reactor is well guarded against all possible instrument failures and human errors. This report presents the basic theory of operation of the two designs alongwith a comparative safety analysis. (auth.)

Design of JRR-4 operation Supporting System

International Nuclear Information System (INIS)

Takahashi, Hiroki; Arai, Nobuyoshi; Yamamoto, Kazuyoshi; Yoritsune, Tsutomu

1999-01-01

A work on the Japan Research Reactor No. 4 (JRR-4) core modification from high-enriched fuels to low-enriched ones was started in 1996 and completed in May 1998. As a part of the work, Operation Supporting System (OSS) was adopted for the purpose of reducing load of operator, preventing operator's error and supplying operation data to users. As the result of adopting this system, integrated monitoring, automation of part of reactor operation and data management were realized. In addition, the cooling control system was compacted. (author)

Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

International Nuclear Information System (INIS)

2002-02-01

The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

Medical set-up of boron neutron capture therapy (BNCT) for malignant glioma at the Japan research reactor (JRR)-4

International Nuclear Information System (INIS)

Yamamoto, T.; Matsumura, A.; Nose, T.; Shibata, Y.; Nakai, K.; Sakurai, F.; Kishi, T.; Kumada, H.; Yamamoto, K.; Torii, Y.

2001-01-01

The University of Tsukuba project for boron neutron capture therapy (BNCT) was initiated at the Japan Atomic Energy Research Institute (JAERI) in 1992. The clinical study for BNCT began at the Japan Research Reactor (JRR)-2 of the JAERI in November 1995. By the end of 1998, a new medical irradiation facility had been installed in JRR-4 of that included a new medical treatment room and patient-monitoring area adjacent to the irradiation room. The medical treatment room was built to reflect a hospital-type operation room that includes an operating table with a carbon head frame, anesthesia apparatus with several cardiopulmonary monitors, etc. Following craniotomy in the treatment room, a patient under anesthesia is transported into the irradiation room for BNCT. The boron concentration in tissue is measured with prompt gamma ray analysis (PGA) and simultaneously by inductively coupled plasma atomic emission spectroscopy (ICP-AES) methods. For the immediate pre- and post-BNCT care, a collaborating neurosurgical department of the University of Tsukuba was prepared in the vicinity of the JAERI. The long term follow-up is done at the University of Tsukuba Hospital. Epithermal neutron beam also became available at the new JRR-4. By changing the thickness and/or the configuration of heavy water, a cadmium plate, and a graphite reflector, the JRR-4 provides a variety of neutron beams, including three typical beams (Epithermal mode and Thermal modes I and II). Intraoperative BNCT using the thermal beam is planned to study at the beginning of the clinical trial. The ongoing development of the JAERI Computational Dosimetry System (JCDS) and radiobiological studies have focused in the application of the epithermal beam for BNCT. After obtaining these basic data, we are planning to use the epithermal beam for intraoperative BNCT. (author)

Development of breast cancer irradiation technique for BNCT at JRR-4

International Nuclear Information System (INIS)

Nakamura, Takemi; Horiguchi, Hironori; Arai, Masaji; Yanagie, Hironobu

2014-06-01

In the Department of Research Reactor and Tandem Accelerator, developments of irradiation technique with application enlargement for breast cancer on BNCT have been performed in the second medium term plans. We compiled this report about the technological development to solve several problems with the irradiation of breast cancer in the medical irradiation facility of JRR-4. In the present study, design fabrication of a collimator for breast cancer, dose evaluation analysis by clinical model, investigation of dose enhancement at deeper region and investigation of fixing method for breast cancer irradiation were studied. By these evaluation results, we verified that the developed breast cancer irradiation technique can be applied to BNCT medical irradiation of JRR-4. These results are expected to be able to contribute to breast cancer irradiation techniques of other reactor-based BNCT and future accelerator-based BNCT. (author)

Maintenance management during long-term stop of JRR-3

International Nuclear Information System (INIS)

Matsui, Yutaka; Suwa, Masayuki; Wada, Shigeru

2016-01-01

JRR-3 has been continuing its stop state for more than 5 years as of FY2015 after stopping the use operation of the reactor in November 2010. Different responses are required in the maintenance management for operation resumption, compared with those in normal operation. This paper introduces part of the maintenance managements that are performed during long-term stop. The water qualities of primary coolant and secondary coolant are controlled by measuring pH and conductivity, and the prevention of crevice corrosion of equipment is performed. In the management of pumps for coolant circulation, vibration measurement is performed to confirm that there are no signs of abnormality. As the management of the core structure, the bend measurement of beryllium reflector is performed to confirm that there is no hindrance to operation resumption, and the visual inspection of high burnup fuel elements is performed to confirm that abnormality has not occurred. As for the management of monitoring equipment, the equipment required in shutdown period is subjected to calibration work, and the equipment required in operation period is subjected to soundness check based on the results of cooling system operation. As the functional maintenance of the cooling system, cooling system operating test and 10-day continuous operation of the cooling system are monthly performed to confirm the soundness of equipment. In addition, the competence maintenance and capacity improvement of operators are performed through the cooling system operation and reactor simulator training. (A.O.)

Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-02-01

The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

JRR-3 cold neutron source facility H2-O2 explosion safety proof testing

International Nuclear Information System (INIS)

Hibi, T.; Fuse, H.; Takahashi, H.; Akutsu, C.; Kumai, T.; Kawabata, Y.

1990-01-01

A cold Neutron Source (CNS) will be installed in Japan Research Reactor-3 (JRR-3) in Japan Atomic Energy Research Institute (JAERI) during its remodeling project. This CNS holds liquid hydrogen at a temperature of about 20 K as a cold neutron source moderator in the heavy water area of the reactor to moderate thermal neutrons from the reactor to cold neutrons of about 5 meV energy. In the hydrogen circuit of the CNS safety measures are taken to prevent oxygen/hydrogen reaction (H 2 -O 2 explosion). It is also designed in such manner that, should an H 2 -O 2 explosion take place, the soundness of all the components can be maintained so as not to harm the reactor safety. A test hydrogen circuit identical to that of the CNS (real components designed by TECHNICATOME of France) was manufactured to conduct the H 2 -O 2 explosion test. In this test, the detonation that is the severest phenomenon of the oxygen/hydrogen reaction took place in the test hydrogen circuit to measure the exerted pressure on the components and their strain, deformation, leakage, cracking, etc. Based on the results of this measurement, the structural strength of the test hydrogen circuit was analyzed. The results of this test show that the hydrogen circuit components have sufficient structural strength to withstand an oxygen/hydrogen reaction

Annual report of Department of Research Reactor, 1997. April 1, 1997 - March 31, 1998

International Nuclear Information System (INIS)

1999-01-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of the department in fiscal year of 1997 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

Annual report of department of research reactor, 1995 (April 1, 1995 - March 31, 1996)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1995 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

Annual report of Department of Research Reactor, 1996. April 1, 1996 - March 31, 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1996 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

Calculation using MVP and MVP-BURN in JRR-3

International Nuclear Information System (INIS)

Komeda, Masao; Kato, Tomoaki; Murayama, Yoji; Yamashita, Kiyonobu

2007-01-01

MVP is the particle-transport Monte Carlo code that has been developed in JAEA. MVP-BURN is an added function to do burn-up calculation. It is easy to built complex structure like core for MVP. And it is easy to do calculations of keff, any reaction rate, flux, burn-up and so on. In this report, it is introduced MVP and MVP-BURN. And some sample calculations of JRR-3 are shown. (author)

Clearance of concrete debris generated from modification work of JRR-3

International Nuclear Information System (INIS)

Satoyama, Tomonori; Nanri, Tomohiro; Kishimoto, Katsumi

2014-01-01

The Japan Atomic Energy Agency (JAEA) planned to apply the clearance system to slightly contaminated concrete debris with radionuclides, which was generated from the modification of the Japan Research Reactor No. 3 (JRR-3) in the Nuclear Science Research Institute (NSRI) of JAEA. The modification work was conducted from 1985 to 1990 and the generated concrete debris has been stored as radioactive waste in interim storage facilities in the NSRI. This is the Japanese first approved of clearance system to concrete debris and stored waste. We established procedures for measuring and evaluating the radioactivity concentration of scored concrete debris. In 2008, the authority approved these procedures for clearance. Since 2009, we have been measuring and evaluating the radioactivity concentration of concrete debris, using the approved procedures. By the end of 2012, the authority had confirmed the correctness of our measurement and evaluation results on about 2,600 tons of concrete debris. About 1,800 tons of those cleared concrete were recycled to flat surface subsidence in the NSRI caused by the Great East Japan Earthquake, and also used as base material to construct new buildings and parking lots in the NSRI. (author)

Nuclear data usage for research reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

1996-01-01

In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

Application of JAERI research reactors to education

International Nuclear Information System (INIS)

Ogawa, Shigeru; Morozumi, Minoru

1987-01-01

At the dawning of the atomic age in Japan, training on reactor operation and reactor engineering experiments has been started in 1958 using JRR-1 (a 50 kW water boiler type reactor with liquid fuel), which was the first research reactor in Japan. The role of the training has been transferred to JRR-4 (a 3500 kW swimming pool type reactor with ETR type fuel) since 1969 due to the decommission of JRR-1. The training courses which have been held are: JRR-1 Short-Term Course for Operation (1958 ∼ 1963) General Course (1961 ∼ ) Reactor Engineering Course (1976 ∼ ) Training Course in Nuclear Technology (International course)(1986 ∼ ). And individual training concerning research reactors for the participants of scientist exchange program sponsored by Science and Technology Agency and of bilateral agreement have been initiated in 1985. The graduates of these courses work as staff members in various fields in nuclear industry. (author)

The heavy water accountancy for research reactors in JAERI

International Nuclear Information System (INIS)

Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

1998-11-01

The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

Proceedings of the first symposium on utilization of research reactors and JMTR

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

The first symposium on utilization of research reactors (JRR-2, JRR-3M, JRR-4) and Japan Materials Testing Reactor (JMTR) in JAERI was held from September 29th to 30th, 1997 at Sannomaru Hotel, Mito. The purpose of this symposium is to announce contribution to progress of scientific technology as well as to promote future utilization of the research reactors and JMTR. During the symposium, 16 reports were presented on nuclear fuel and material, neutron beam experiment, medical irradiation, radioisotope production and neutron activation analysis. The present status of the research reactors and JMTR were also reported. The special lecture titled `JRR-2 and Medical Irradiation` was given by Mr. Nakamura, former editorial writer of Yomiuri. Finally, panel discussion was carried on `The Role of Research Reactors and JMTR in Scientific Technology for the future` actively by the participants and experts in every field of research reactor utilization. 250 people participated in this symposium from universities, national research institutes, private corporations and JAERI. This proceedings briefly summarizes 16 reports, the content of panel discussion and so forth. (J.P.N.)

Annual report of department of research reactor, 1999. April 1, 1999 - March 31, 2000

International Nuclear Information System (INIS)

2001-03-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3M (new JRR-3) and the JRR-4 and for the related R and D. Besides the decommissioning of the JRR-2 and RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 1999 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

Research report on the users' needs for next research reactor

International Nuclear Information System (INIS)

Takahashi, Hiroyuki; Tamura, Itaru; Hosoya, Toshiaki; Horiguchi, Hironori

2015-03-01

JRR-3 has been operated for more than 25 years for that it is time to investigate the role of a next research reactor. A task force under the Committee for Promotion of JRR-3 Neutron Beam Application has been organized by Department of Research Reactor and Tandem Accelerator to survey neutron beam application trends in the future. This is a report on the survey results and users' requirements for the next research reactor have been summarized in this report carried by the task force. (author)

Present status and future perspective of research and test reactors in JAERI

International Nuclear Information System (INIS)

Baba, Osamu; Kaieda, Keisuke

1999-01-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

Present status and future perspective of research and test reactors in JAERI

Energy Technology Data Exchange (ETDEWEB)

Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

1999-08-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

Thermohydraulic characteristics analysis of natural convective cooling mode on the steady state condition of upgraded JRR-3 core, using COOLOD-N code

International Nuclear Information System (INIS)

Kaminaga, Masanori; Watanabe, Shukichi; Ando, Hiroei; Sudo, Yukio; Ikawa, Hiromasa.

1987-03-01

This report describes the results of the steady state thermohydraulic analysis of upgraded JRR-3 core under natural convective cooling mode, using COOLOD-N code. In the code, function to calculate flow-rate under natural convective cooling mode, and a heat transfer package have been newly added to the COOLOD code which has been developed in JAERI. And this report describes outline of the COOLOD-N code. The results of analysis show that the thermohydraulics of upgraded JRR-3 core, under natural convective cooling mode have enough margine to ONB temperature, DNB heat flux and occurance of blisters in fuel meats, which are design criterion of upgraded JRR-3. (author)

Achievement and development of neutron beam utilization in research reactors

International Nuclear Information System (INIS)

Isshiki, Masahiko

1996-01-01

Especially regarding the neutron beam experiment in Japan, the basic research has been developed by utilizing the JRR-2 of Japan Atomic Energy Research Institute and the KUR of Kyoto University over long years. Now, the JRR-3M of JAERI was revived as a high performance, general purpose reactor, and bears important roles as the neutron beam experiment center in Japan. Thanks to one of the most powerful reactor neutron sources in the world and the cold neutron source, the environment of research was greatly improved, and the excellent results of researches began to be reported. The discovery of neutrons by Chadwick and the history of the related researches are described. As neutron sources, radioisotopes, accelerators and nuclear reactors are properly used corresponding to purposes. As the utilization of research reactors for neutron sources, the utilization for irradiation and neutron beam experiment are carried out. The outline of the research reactor JRR-3M is explained. The state of utilization in neutron scattering experiment, neutron radiography, prompt γ-ray analysis and the medical irradiation of neutrons is reported. (K.I.)

The boron concentration measurement by the prompt gamma-ray analysis device at JRR-4

International Nuclear Information System (INIS)

Hori, N.; Torii, Y.; Yamamoto, K.; Kishi, T.; Takada, J.

2000-01-01

In determining the dose for BNCT, it is important to measure the boron concentration in blood as well as the thermal neutron fluence. To attain an accurate measurement of the boron concentration in the whole blood at about 30 ppm, the Prompt Gamma-ray Analysis (PGA) is adopted to enable the measurement with a high degree of accuracy in short time, since the chemical analysis technique has some difficulties on it. In order to install the PGA device in a swimming pool type research reactor JRR-4, the super-mirror for obtaining the necessary neutron flux in the space over the pool surface was perpendicularly placed, because the reactor had no horizontal experiment tube for this kind of use. As the accurate measurement of the boron concentration in whole blood using PGA is enabled herewith, the first medical irradiation for BNCT at JRR-4 was carried out on October 25, 1999. (author)

Characteristic measurements of JRR-2 after modification

International Nuclear Information System (INIS)

1977-03-01

JRR-2 was shut down in December 1973, to improve it in heavy water leakage from the metal packing between core tank and support ring, corrosion of the lower shielding plug, and fault in the control-rod mechanism. Modification was on a standing seal to stop heavy water leakage at the support ring, replacement of the reactor top shield, replacement of the control-rod mechanism, improvement of the helium system, improvement of the irradiated air system to reduce radioactive argon gas release from the stack. Works were completed in September 1975. Functional tests and characteristic measurements then followed, to show good results of the modification. Described are; modification, critical experiments, calibration of the control rods, reactor thermal power, measurements of the neutron flux distribution, core heavy water flow distribution, heavy water decomposition rate, helium leak rate and tritium concentration in the thermal shielding light water, radioactive argon gas release from the stack, and effect of the reactor top shield. (auth.)

Annual report on operation, utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1990-09-01

This report describes the activities of the Department of Research Reactor Operation in fiscal year of 1989. It also presents some technical topics on the reactor operation and utilization in details. The Department is responsible for operation of the research reactors, JRR-2 and JRR-4, and the Hot Laboratory. The research reactor JRR-3 was reconstructed to enhance the performance for utilization. The first criticality was achieved on March 22, 1989, and it subsequently went into operation. In connection with the reactor operation, the various research and development activities in the area of fuel management, water chemistry, radiation monitoring and material irradiation have been made. In the Hot Laboratory, post-irradiation examinations of fuels and materials have been carried out along with the development of related techniques. (author)

Medical setup of intraoperative BNCT at JRR-4

International Nuclear Information System (INIS)

Akutsu, H.; Yamamoto, T.; Matsumura, A.

2000-01-01

Since October 1999, we have been performing clinical trials of intraoperative boron neutron capture therapy (IOBNCT) using a mixed thermal-epithermal beam at the Japan Research Reactor No. 4 (JRR-4). For immediate pre-BNCT care, including administration of a boron compound as well as post-BNCT care, a collaborating neurosurgical department of the University of Tsukuba was prepared in the vicinity of JRR-4. Following craniotomy in the treatment room, anesthetized patients were transported into the irradiation room for BNCT. The boron concentration in tissue was measured by the PGA and ICP-AES methods. The long-term follow-up was done at the University of Tsukuba Hospital. IOBNCT is a complex clinical procedure, which requires sophisticated operating team and co-medical staffs and also cooperation with physicist team. IOBNCT is a complex clinical procedure requiring a high level of cooperation among the operating team, co-medical staff, and physicists. For the safe and successful performance of IOBNCT, we have made the program including critical pathway and prepared various equipments for IOBNCT. To ensure the safe and successful performance of IOBNCT, we developed a critical pathway for use during the procedure, and prepared various apparatus for IOBNCT. (author)

Neutron spin echo spectrometer at JRR-3M

International Nuclear Information System (INIS)

Takeda, Takayoshi; Komura, Shigehiro; Seto, Hideki; Nagai, Michihiro; Kobayashi, Hideki; Yokoi, Eiji; Ebisawa, Tooru; Tasaki, Seiji.

1993-01-01

We have designed and have been constructing at C 2-2 cold neutron guide port of JRR-3M, JAERI, a neutron spin echo spectrometer (NSE) which is equipped with two optimized magnets for neutron spin precession, a position sensitive detector (PSD), a converging polarizer and a wide area analyzer. The dynamic range of scattering vector Q covers from 0.01 A -1 to 0.3 A -1 and that of energy E from 30neV to 0.1meV. This spectrometer makes it possible to study a mesoscopic spatial structure of the order of 1-100nm combined with a nanosecond temporal structure of the order of 0.1-100ns corresponding to dynamical behavior of large molecules such as polymer. A test experiment shows that the homogeneity condition of the precession magnet is loosened by means of PSD. (author)

Report on operation, utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1980-03-01

Activities of the Division of Research Reactor Operation in fiscal 1978 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

Report on operation utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1982-03-01

Activities of the Division of Research Reactor Operation in fiscal 1980 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

Report on operation, utilization and technical development of Research Reactors and Hot Laboratory

International Nuclear Information System (INIS)

1984-10-01

Activities of the Division of Research Reactor Operation in fiscal 1981 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

Current situation and future of research reactor in Japan

International Nuclear Information System (INIS)

Metoki, Naoto

2011-01-01

A new style of neutron facilities coexisting high intensity pulse neutron source (J-PARC/MLF) and steady state neutron from research reactor (JRR-3) is proposed. The coexistence of the pulse and reactor source is the world trend which is inevitable for the next generation neutron science. A sophisticated management and comprehensive user program are important for Japanese neutron community, involving high performance instruments with pulse neutron and steady neutron source with widely spread users. JAEA is the most responsible for the future of neutron science, because of the commitment for both type of neutron source, J-PARC/MLF and JRR-3. (author)

Present status and future perspectives of research and test reactor in Japan

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Kaieda, Keisuke

2000-01-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

Present status and future perspectives of research and test reactor in Japan

Energy Technology Data Exchange (ETDEWEB)

Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2000-10-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

Analytical study on thermal-hydraulic behavior of transient from forced circulation to natural circulation in JRR-3

International Nuclear Information System (INIS)

Hirano, Masashi; Sudo, Yukio

1986-01-01

Transient thermal-hydraulic behaviors of the JRR-3 which is an open-pool type research reactor has been analyzed with the THYDE-P1 code. The focal point is the thermal-hydraulic behaviors related to the core flow reversal during the transition from forced circulation downflow to natural circulation upflow. In the case of a loss-of-coolant accident (LOCA), for example, the core flow reversal is expected to occur just after the water pool isolation from the primary cooling loop with a leak. The core flow reversal should cause a sudden increase in fuel temperature and a steep decrease in the departure-from-nucleate-boiling ratio (DNBR) and the phenomenon is, therefore, very important especially for safety design and evaluation of research reactors. Major purposes of the present work are to clarify physical phenomena during the transient and to identify important parameters affecting the peak fuel temperature and the minimum DNBR. The results calculated with THYDE-P1 assuming the sequences of events of the loss-of-offsite power and LOCA help us to understand the phenomena both qualitatively and quantitatively, with respect to the safety design and evaluation. (author)

Activity report on the utilization of research reactors. Japanese fiscal year, 2002

International Nuclear Information System (INIS)

2004-08-01

During the fiscal year 2002, the Tokai Research Establishment research reactors carried out 7 cycles of joint use reactor operation at JRR-3 and 39 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, Irradiation Test of Reactor Materials and fission track. This volume contains 279 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analysis, reactor materials, prompt gamma analysis, and others, submitted by the users in JAERI and from other organizations. (author)

Activity report on the utilization of research reactors. Japanese fiscal year, 2003

International Nuclear Information System (INIS)

2005-09-01

During the fiscal year 2003, the Tokai Research Establishment research reactors carried out 8 cycles of joint use reactor operation at JRR-3 and 42 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, Irradiation Test of Reactor Materials and fission track. This volume contains 246 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analysis, reactor materials, prompt analysis, and others, submitted by the users in JAERI and from other organizations. (author)

Annual report on operation, utilization and technical development of research reactors and hot laboratory, from April 1, 1987 to March 31, 1988

International Nuclear Information System (INIS)

1988-10-01

Activities of the Department of Research Reactor Operation in fiscal year 1987 are described. The department is responsible for operation and maintenance of JRR-2, JRR-4, Research Reactor Development Division which performed upgraded JRR-3 and other R D, and Hot Laboratory. In the above connection various other work has also been performed, such as technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, we have performed post-irradiation examinations of fuels and materials, and also development of examination procedures, too. (author)

Annual report on operation, utilization and technical development of research reactors and hot laboratory, from April 1, 1985 to March 31, 1986

International Nuclear Information System (INIS)

1986-10-01

Activities of the Department of Research Reactor Operation in fiscal year 1985 are described. The department is responsible for operation and maintenance of JRR-2, JRR-4, Research Reactor Development Division which performed upgraded JRR-3 and other R and D, and Hot Laboratory. In the above connection various other work has also been performed, such as technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, we have performed post-irradiation examinations of fuels and materials, and also development of examination procedures, too. (author)

Activity report on the utilization of research reactors. Japanese fiscal year, 2004

International Nuclear Information System (INIS)

2006-03-01

During the fiscal year 2004, the Nuclear Science Research Institute research reactors carried out 7 cycles of joint use reactor operation at JRR-3 and 41 cycles at JRR-4. The research reactors are being utilized for various purposes including experimental studies such as neutron scattering, prompt gamma analysis, neutron radiography and medical irradiation (BNCT), and irradiation utilization such as neutron activation analysis of various samples, RI production, Irradiation Test of Reactor Materials and fission track, advanced Science Research. This volume contains 235 activity reports, which are categorized into the fields of neutron scattering (10 subcategories), neutron radiography, neutron activation analysis, RI Production, prompt gamma analysis, and others, submitted by the users in JAEA and from other organizations. (author)

Status of reduced enrichment programs for research reactors in Japan

International Nuclear Information System (INIS)

Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo

1997-01-01

The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE

Status of reduced enrichment programs for research reactors in Japan

Energy Technology Data Exchange (ETDEWEB)

Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

1997-08-01

The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

Experiments of JRR-4 low-enriched-uranium-silicied fuel core

International Nuclear Information System (INIS)

Hirane, Nobuhiko; Ishikuro, Yasuhiro; Nagadomi, Hideki; Yokoo, Kenji; Horiguchi, Hironori; Nemoto, Takumi; Yamamoto, Kazuyoshi; Yagi, Masahiro; Arai, Nobuyoshi; Watanabe, Shukichi; Kashima, Yoichi

2006-03-01

JRR-4, a light-water-moderated and cooled, swimming pool type research reactor using high-enriched uranium plate-type fuels had been operated from 1965 to 1996. In order to convert to low-enriched-uranium-silicied fuels, modification work had been carried out for 2 years, from 1996 to 1998. After the modification, start-up experiments were carried out to obtain characteristics of the low-enriched-uranium-silicied fuel core. The measured excess reactivity, reactor shutdown margin and the maximum reactivity addition rate satisfied the nuclear limitation of the safety report for licensing. It was confirmed that conversion to low-enriched-uranium-silicied fuels was carried out properly. Besides, the necessary data for reactor operation were obtained, such as nuclear, thermal hydraulic and reactor control characteristics. This report describes the results of start-up experiments and burnup experiments. The first criticality of low-enriched-uranium-silicied core was achieved on 14th July 1998, and the operation for joint-use has been carried out since 6th October 1998. (author)

Construction of JRR-3 spent fuel dry storage facility

International Nuclear Information System (INIS)

Adachi, M.

1982-01-01

To store the JRR-3 metallic natural uranium spent fuel elements, dry storage facility has been constructed in JAERI. This facility has a capacity of about 30T of uranium. The elements are placed in encapsulated canister, then stored in drywell in the store. The store is basically an ordinary concrete box, about 12m long, 13m wide, and 5m deep. The store comprises a 10 x 10 lattice array of the drywells. The drywell consists of a stainless steel liner which is 2.5m deep, 36cm ID and 0.8cm thickness. A drywell also has an air inlet, outlet pipe for radiation monitoring and a shield plug in carbon steel for radiation protection. A canister which consists of stainless steel with 0.5cm thickness contains 36 elements. Sealing of the canister is accomplished by fusion welding

Replacement of the moderator cell unit of JRR-3's cold neutron source facility

International Nuclear Information System (INIS)

Hazawa, Tomoya; Nagahori, Kazuhisa; Kusunoki, Tsuyoshi

2006-10-01

The moderator cell of the JRR-3's cold neutron source (CNS) facility, converts thermal neutrons into cold neutrons by passing through liquid cold hydrogen. The cold neutrons are used for material and life science research such as the neutron scattering. The CNS has been operated since the start of JRR-3's in 1990. The moderator cell containing liquid hydrogen is made of stainless steel. The material irradiation lifetime is limited to 7 years due to irradiation brittleness. The first replacement was done by using a spare part made in France. This replacement work of 2006 was carried out by using the domestic moderator cell unit. The following technologies were developed for the moderator cell unit production. 1) Technical development of black treatment on moderator cell surface to increase radiation heat. 2) Development of bending technology of concentric triple tubes consisting from inside tube, Outside tube and Vacuum insulation tube. 3) Development of manufacturing technique of the moderator cell with complicated shapes. According to detail planed work procedures, replacement work was carried out. As results, the working days were reduced to 80% of old ones. The radiation dose was also reduced due to reduction of working days. It was verified by measurement of neutrons characteristics that the replaced moderator cell has the same performance as that of the old moderator cell. The domestic manufacturing of the moderator cell was succeeded. As results, the replacement cost was reduced by development of domestic production technology. (author)

Annual report on operation, utilization and technical development of Research Reactors and Hot Laboratory, from April 1, 1983 to March 31, 1984

International Nuclear Information System (INIS)

1984-11-01

Activities of the Department of Research Reactor Operation in fiscal year 1983 are described. The department is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other work has also been performed, such as technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, we have performed post-irradiation examinations of fuels and materials, and also development of examination procedures, too. (author)

Status of reduced enrichment program for research reactors in Japan

International Nuclear Information System (INIS)

Kaieda, Keisuke; Baba, Osamu; Nagaoka, Yoshiharu; Kanda, Keiji; Nakagome, Yoshihiro

1999-01-01

The reduced enrichment programs for the JRR-3M, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI) have been completed. The KUR of Kyoto University Research Reactor Institute (KURRI) has been partially completed and is still in progress under the Joint Study Program with Argonne National Laboratory (ANL). The JRR-3M commenced using LEU silicide fuel elements instead of LEU aluminide fuel elements in September, 1999. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and April 1994 the U.S. Government gave an approval to utilize HEU fuel in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until March 2004, then the full core conversion with LEU silicide will be done. The first shipment of spent fuels since 1974 was done in August, 1999. (author)

JRR-3を用いた99Mo製造に関する概念検討

OpenAIRE

広瀬 彰; 米田 政夫; 木名瀬 政美; 反田 孝美; 和田 茂

2010-01-01

放射性医薬品として核医学の分野で最も多く用いられているテクネチウム-99m(99mTc)の親核種であるモリブデン-99(99Mo)について、JRR-3を用いた製造を検討する。99Moは、その供給をすべて輸入に依存している状況にあることから、産業界と共同で99Moの一部国産化を目指すものである。本報告書では、JRR-3研究用原子炉を用いた(n,gamma)反応を利用し99Moを得る「中性子放射化法」による99Moの製造プロセスの技術的な検討について述べる。...

Annual report of department of research reactors, 2001. April 1, 2001 - March 31, 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-12-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2001 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

Annual report of Department of Research Reactor, 2003. April 1, 2003 - March 31, 2004

International Nuclear Information System (INIS)

2005-02-01

The Department of Research Reactor is responsible for the operation, maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides the RI production and its R and D are carried out. This report describes the activities of the department in fiscal year of 2003 and also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, utilization of irradiation and neutron beam experiments, technical management including management of fuels and water chemistry, and related R and D works. The RI production and its R and D works were conducted as well. The international co-operations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. (author)

Investigation of the crack in a reflector element of JRR-4

International Nuclear Information System (INIS)

2008-09-01

The JRR-4 (Japan Research Reactor No.4) has been used as a shared utilization facility for many researches since the first criticality in January 1965. A crack was ascertained on a weld area of one reflector element on December 28, 2007. The Department of Research Reactor and Tandem Accelerator set up an ad hoc working group of experts in the JAEA (Japan Atomic Energy Agency), and investigated cause of crack on the weld area. The following examinations were carried out; visual examination, dimensional examination, fractography examination and so on. It was concluded that the main cause of the crack is the swelling of graphite in the reflector element. The swelling must be due to neutron irradiation. We carried out a radiografical examination of the other reflector elements. As the result, we determined that many of them were not in a suitable state to be used because of swelling of graphite. The design of the new reflector elements should be carried out, based on the relation between the irradiation does and swelling rate, which has been obtained in these investigation. (author)

Neutronics comparative analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON and DONJON are applied and verified in calculations of research reactors. • Continuous-energy Monte Carlo calculations by RMC are chosen as the references. • “ECCO” option of DRAGON is suitable for the calculations of research reactors. • Manual modifications of cross-sections are not necessary with DRAGON and DONJON. • DRAGON and DONJON agree well with RMC if appropriate treatments are applied. - Abstract: Simulation of the behavior of the plate-type research reactors such as JRR-3M and CARR poses a challenge for traditional neutronics calculation tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity and large leakage of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON and DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic approach. The goal of this research is to examine the capability of the deterministic code system DRAGON and DONJON to reliably simulate the research reactors. The results indicate that the DRAGON and DONJON code system agrees well with the continuous-energy Monte Carlo simulation on both k eff and flux distributions if the appropriate treatments (such as the ECCO option) are applied

Verification of vectorized Monte Carlo code MVP using JRR-4 experiment of fast neutrons penetrating through graphite and water

International Nuclear Information System (INIS)

Odano, N.; Miura, T.; Yamaji, A.

1996-01-01

Measurement of activation reaction rates was carried out for fast neutrons penetrating through graphite and water from the core of JRR-4 research reactor of JAERI, with paying attention to the energy above 10 MeV. Analysis of the experiment was made using a vectorized continuous energy Monte Carlo code MVP to verify the code. The analysis shows good agreements between the measurement and calculation and the MVP code has been confirmed its validity for the fast neutron transport calculations above 10 MeV in fission neutron field. (author)

Application of MCNP in the criticality calculation for reactors

International Nuclear Information System (INIS)

Zhong Zhaopeng; Shi Gong; Hu Yongming

2003-01-01

The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

State of maintenance of JRR-3M neutron radiography facilities

International Nuclear Information System (INIS)

Tsuruno, Akira

1994-01-01

Nearly two years have elapsed since the use of the JRR-3M neutron radiography facilities was begun. Five years passed since the decision of the detailed specification, and this period is enough for the contained equipment and the peripheral equipment become old-fashioned. In addition, as unexpected high performance was obtained by these facilities, the measures were to be taken in many places. In this fiscal year, the improvement of the contained equipment and the increase of the peripheral equipment were carried out. These improvement and renewal are still insufficient for these facilities demonstrate the performance fully, and are expected to continue hereafter. As the measures taken in relation to the TNRF-2, the renewal of the BN shutter, the remodeling of the black box and the improvement of the interlock were carried out. As the measures taken in relation to CNRF, the new installation of auxiliary shields, the remodeling of the shutter and the new installation of internal illumination were carried out. As the measures taken in relation to the image processor, the outline based on long term perspective is shown. The equipments which were purchased in this fiscal year are the work station which becomes the center of the image-processing network, the cooled CCD camera system and the SIT camera for renewal. In the dark room, a scanning micro-densitometer was installed. (K.I.)

Evaluation of JRR-4 neutron beam using tumor cells

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, Kazuyoshi; Kumada, Hiroaki; Torii, Yoshiya; Kishi, Toshiaki; Horiguchi, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yamamoto, Tetsuya; Matsumura, Akira; Nose, Tadao [Tsukuba Univ., Ibaraki (Japan)

2001-03-01

For preparation of irradiation plan of boron-neutron capture therapy (BNCT), not only the physical dose is important, but also weighted factors or RBE are also necessary on the evaluation of the effect on the organism. Physical dose calculated by dose evaluation system (JCDS : JAERI Computational Dosimetry System) must appropriately carry out the weighting by various cells like tumor, central nerve, glia, and the vascular in proportion to JRR-4 each irradiation mode. In-vitro biological experiment which used 9L gliosarcoma and C6 glioma in the head water phantom was carried out in order to evaluate these effect. Neutron beam characteristics of JRR-4 were also evaluated from the functions of survival fraction of these cells. As a result of the evaluation, it became clear that the dose evaluation calculated from physical dose of the boron and nitrogen carried out in traditional BNCT of Japan using thermal neutron is applicable for thermal and epi-thermal mixed neutron beam. (author)

COOLOD, Steady-State Thermal Hydraulics of Research Reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-01-01

1 - Description of program or function: The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is a revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode. A 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is a subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. 2 - Method of solution: The 'Heat Transfer Package' is a subprogram for calculating heat transfer coefficients, ONB temperature, heat flux at onset of flow instability and DNB heat flux. The 'Heat transfer package' was especially developed for research reactors which are operated under low pressure and low temperature conditions using plate-type fuel, just like the JRR-3M. Heat transfer correlations adopted in the 'Heat Transfer Package' were obtained or estimated based on the heat transfer experiments in which thermal-hydraulic features of the upgraded JRR-3 core were properly reflected. The 'Heat Transfer Package' is applicable to upward and downward flow

Russian Translations of Writings of J.R.R. Tolkien as the Stage of their Cultural Reception in Russia

Directory of Open Access Journals (Sweden)

Ellina V. Shustova

2017-11-01

Full Text Available This article identifies problems of translation of J.R.R. Tolkien`s works to the Russian language. Due to complexity of their reception in the Russian literature and culture, these problems continue at the present time. The purpose of this article is to consider the translational interpretations as a form of reception of J.R.R. Tolkien`s works in Russia and perform analysis of relevant stages of the receptive process. The leading approach to the study of this problem is the methodology of receptive aesthetics; this takes into account the development trends of the method. This research shows that it is the translation, in many respects, that sets trends for further understanding of writings by J.R.R. Tolkien in Russia and includes them in a Russian cultural and literary context, as well as set their level of systematic perception. Contents of this article may be used for readings in general and special courses for Western literature at universities, and may be useful to a wide range of readers interested in the work of J.R.R. Tolkien.

Mechanical strength evaluation of the glass base material in the JRR-3 neutron guide tube

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tetsuya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-02-01

The lifetime of the thermal neutron guide tube installed JRR-3 was investigated after 6 years from their first installation. And it was confirmed that a crack had been piercing into the glass base material of the side plate of the neutron guide tube. The cause of the crack was estimated as a static fatigue of the guide tube where an inside of the tube had been evacuated and stressed as well as an embrittlement of the glass base material by gamma ray irradiation. In this report, we evaluate the mechanical strength of the glass base material and estimate the time when the base material gets fatigue fracture. Furthermore, we evaluate a lifetime of the neutron guide tube and confirm the validity of update timing in 2000 and 2001 when the thermal neutron guide tubes T1 and T2 were exchanged into those using the super mirror. (author)

Measurement of tritium activity in the aluminum pipe of JRR-2 heavy water primary cooling system using imaging plate

International Nuclear Information System (INIS)

Motoishi, Shoji; Kobayashi, Katsutoshi

2000-12-01

JRR-2 is the heavy water cooling type nuclear reactor, which has been operated for 36 years (1960-1976) and in the process of decommissioning at present. For this reason, evaluation of tritium quantity permeated into the pipe and apparatus of the primary coolant heavy water circulating system is important. In the Radioisotope Production Division, activity of tritium in aluminum pipe was measured with imaging plate (IP), liquid scintillation analyzer and high purity germanium detector (HPGe). After acrylic paints was applied for the region except for tritium contamination on the surface of aluminum pipe, only the oxidized contaminated part was dissolved by 1.5%(1.21M) HF for 3 minutes, and measured with IP. As a result, the tritium was found to permeate in the depth of 25 μm. Moreover, 90% of it was found to be distributed within 7 μm. (author)

Determination of boron in ceramic reference materials by prompt gamma activation analysis using focused neutron guided beam of JRR-3M

International Nuclear Information System (INIS)

Miura, T.; Kuroiwa, T.; Chiba, K.; Matsue, H.

2008-01-01

Prompt gamma activation analysis using a focused thermal neutron guided beam at JAEA JRR-3M was applied to the determination of B in ceramic certified reference materials (BAM CRM S-003 Silicon Carbide Powder and NMIJ CRM 8004-a Silicon Nitride Powder). Cl and Si were used as internal standards to obtain linear calibration curves of B. The analytical result of B in BAM CRM S-003 was in good agreement with the certified value. The relative expanded measurement uncertainties (k = 2) were 4.8% for BAM CRM S-003 and 4.9% for NMIJ CRM 8004-a. (author)

Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

Activity report on neutron scattering research. V. 1, 1994

International Nuclear Information System (INIS)

Fujii, Y.; Oohara, Y.

1994-09-01

In April, 1993, the Neutron Scattering Laboratory attached to the Institute for Solid State Physics, University of Tokyo, was newly established in Tokai, Ibaraki Prefecture, to promote nationwide users' programs for utilizing the university-owned neutron instruments installed at the JRR-3M reactor of Japan Atomic Energy Research Institute. This upgraded reactor (20 MW, the cold source is installed) has drastically expanded the number of users and research areas since 1990 when it became operational. Currently 8 and 3 out of 18 new spectrometers in total at the JRR-3M are owned by ISSP and Tohoku University, respectively, while the remaining 7 spectrometers belong to JAERI. In addition, 3 conventional spectrometers in the 30 years old JRR-2 reactor (10 MW) have also supported research activities. This is the first issue of 'Activity report on neutron scattering research', and it is to be published annually. In this report, the brief history of neutron scattering research, the users' programs, the committees, the neutron scattering instruments available at the JRR-3M and the JRR-2M, the activity reports on structures and excitation, magnetism, superconductors, liquid and glass, material science, polymers, biology and instrumentation, and publication list are reported. (K.I.)

VHTR-fuel irradiation capsules for VT-1 hole of JRR-2

International Nuclear Information System (INIS)

Kikuchi, Teruo; Kikuchi, Akira; Tobita, Tsutomu; Kashimura, Satoru; Miyasaka, Yasuhiko

1977-02-01

Irradiations of VHTR fuels were made in the VT-1 irradiation hole of JRR-2. Three capsules, VP-1, VP-2 and VP-4, which contained fuel compacts, were irradiated for 300 hr at temperatures of 950 0 , 1370 0 and 1500 0 C up to the estimated burn-ups of 0.74, 0.87 and 0.80%FIMA, respectively. And, to study the amoeba effect of fuel particles, two capsules, VP-3 and VP-5, were irradiated for 300 hr at temperatures of 1650 0 and 1670 0 C up to the estimated burn-ups of 0.38 and 0.33%FIMA, respectively. (auth.)

Calculation of tritium release from reactor's stack

International Nuclear Information System (INIS)

Akhadi, M.

1996-01-01

Method for calculation of tritium release from nuclear to environment has been discussed. Part of gas effluent contain tritium in form of HTO vapor released from reactor's stack was sampled using silica-gel. The silica-gel was put in the water to withdraw HTO vapor absorbed by silica-gel. Tritium concentration in the water was measured by liquid scintillation counter of Aloka LSC-703. Tritium concentration in the gas effluent and total release of tritium from reactor's stack during certain interval time were calculated using simple mathematic formula. This method has examined for calculation of tritium release from JRR-3M's stack of JAERI, Japan. From the calculation it was obtained the value of tritium release as much as 4.63 x 10 11 Bq during one month. (author)

Prompt gamma-ray analysis using JRR-3M cold and thermal neutron guide beams

International Nuclear Information System (INIS)

Yonezawa, C.; Haji Wood, A.K.; Magara, M.; Hoshi, M.; Tachikawa, E.; Sawahata, H.; Ito, Y.

1993-01-01

A permanent and stand-alone neutron-induced prompt gamma-ray analysis (PGA) system, usable at both cold and thermal neutron beam guides of JRR-3M has been constructed. Neutron flux at the sample positions were 1.4x10 8 and 2.4x10 7 n cm -2 s -1 for the cold and thermal neutrons, respectively. The γ-ray spectrometer is equipped to acquire three modes of spectra simultaneously: single mode, Compton suppression mode and pair mode, in an energy range up to 12 MeV. Owing to the cold neutron guide beam and the low γ-ray background system, analytical sensitivities and detection limits better than those in other PGA systems have been achieved. Analytical sensitivity and detection limit for 73 elements were measured. Boron, Gd, Sm and Cd are the most sensitive elements with detection limits down to 1 to 10 ng. For some elements such as F, Al, V, Eu and Hf, decay γ-rays are more sensitive compared to their respective prompt γ-ray. Analytical sensitivity of several heavy elements through detection of characteristic X-rays was higher than that through the prompt γ-ray detection. Analytical applicability of some sensitive elements such as B, H, Gd and Sm were examined. Isotopic analysis of Ni and Si were also examined. (author)

J.R.R. Tolkien and Jordanes. Some resemblances in spiritual outlook

Directory of Open Access Journals (Sweden)

Miryam Librán Moreno

2013-12-01

Full Text Available An analysis of three instances of the influence exerted by Jordanes on J.R.R. Tolkien’s legendarium: (a the notion that cultural and political decay may be caused by being exiled from a formerly paradisiacal situation; (b the notion that winning a war against Evil may become an unwanted cause of cultural devolution; (c the construction of a pseudo-history by establishing genetic and cultural links with a mythical, highly evolved people from the legendary past.

Visualization and measurement of pressurized multiphase flow using neutron radiography of JRR-3M system

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yasuo [Yamaguchi Univ. (Japan); Matsubayasi, Masahito

1998-01-01

Concerning the transient phenomenon of solid-gas two-phase flow, an attempt was made to visualize and measure a flow phenomenon in which three-dimensional bubbles occurred, grew and collapsed in the vicinity of a gas injection nozzle while solid particles were circulating. Such a phenomenon could not or hardly be visualized and measured by conventional methods. Such two-phase flow was visualized using neutron radiography, its characteristics measured and the usefulness of the visualization by neutron radiography confirmed. For this purpose, three-dimensional fluidized bed vessels, rectangular or cylindrical-shaped, made of steel or aluminum sheet, were prepared. Polyethylene or glass beads were used as solid particles and activated carbon particles as the tracer. In the experiment, nitrogen gas was blown into the vessel from one nozzle and distributors provided at the bottom of the vessel and exhausted from the top via the exhaust valve, by which the pressure in the vessel was controlled. The imaging was done in the following way: A test chamber was provided beside the vessel to receive neutron beams from the JRR-3M system, the intensity of transmitted neutrons was converted to visible light by scintillator and the images were videotaped. The initial objectives of visualizing and measuring bubbles occurring, growing and collapsing and solid particles circulating in the solid-gas two-phase flow have been achieved by means of neutron radiography. (N.H.)

IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

West, C D [comp.

1990-05-01

Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

West, C.D.

1990-05-01

Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

Precision of neutron activation analysis for environmental biological materials

International Nuclear Information System (INIS)

Hamaguchi, Hiroshi; Iwata, Shiro; Koyama, Mutsuo; Sasajima, Kazuhisa; Numata, Yuichi.

1977-01-01

Between 1973 and 1974 a special committee ''Research on the application of neutron activation analysis to the environmental samples'' had been organized at the Research Reactor Institute, Kyoto University. Eleven research groups composed mainly of the committee members cooperated in the intercomparison programme of the reactor neutron activation analysis of NBS standard reference material, 1571 Orchard Leaves and 1577 Bovine Liver. Five different type of reactors were used for the neutron irradiation; i.e. KUR reactor of the Research Reactor Institute, Kyoto University, TRIGA MARK II reactor of the Institute for Atomic Energy, Rikkyo University, and JRR-2, JRR-3, JRR-4 reactor of Japan Atomic Energy Research Institute. Analyses were performed mainly by instrumental method. Precision of the analysis of 23 elements in Orchard Leaves and 13 elements in Bovine Liver presented by the different research groups was shown in table 4 and 5, respectively. The coefficient of variation for these elements was from several to -- 30 percent. Averages given to these elements agreed well with the NBS certified or reference values. Thus, from the practical point of view for the routine multielement analysis of environmental samples, the validity of the instrumental neutron activation technique for this purpose has been proved. (auth.)

Annual report of the Japan Atomic Energy Research Institute, for fiscal 1987

International Nuclear Information System (INIS)

1988-01-01

Fiscal year 1987 fell on the year to revise the long term program every five years, and the new long term program was decided, and the course that JAERI takes hereafter was shown. In this new program, the safety research and the upgrading research in the development from LWRs to FBRs, the promotion of the research and development of high temperature engineering, nuclear fusion, the utilization of radiation and nuclear-powered ships, the basic research and the development of base technology for opening up atomic energy frontier, and the expansion of international cooperation were emphasized as the role of JAERI. The research on engineering safety, the development of high temperature gas-cooled reactors, the research and development of nuclear fusion, and the research and development of nuclear-powered ships are reported. In the fields of the related research, basic research and the utilization of radiation, the research on the reactor physics of FBRs and thermal neutron reactors, the research on nuclear fuel and reactor materials, the research on reactor dismantling techniques, the development of tritium production technology and many others were advanced. The operation for utilizing JRR 2, JRR 4 and JMTR was carried out, and JRR 3 was under reconstruction. (K.I.)

Reactivity management and burn-up management on JRR-3 silicide-fuel-core

International Nuclear Information System (INIS)

Kato, Tomoaki; Araki, Masaaki; Izumo, Hironobu; Kinase, Masami; Torii, Yoshiya; Murayama, Yoji

2007-08-01

On the conversion from uranium-aluminum-dispersion-type fuel (aluminide fuel) to uranium-silicon-aluminum-dispersion-type fuel (silicide fuel), uranium density was increased from 2.2 to 4.8 g/cm 3 with keeping uranium-235 enrichment of 20%. So, burnable absorbers (cadmium wire) were introduced for decreasing excess reactivity caused by the increasing of uranium density. The burnable absorbers influence reactivity during reactor operation. So, the burning of the burnable absorbers was studied and the influence on reactor operation was made cleared. Furthermore, necessary excess reactivity on beginning of operation cycle and the time limit for restart after unplanned reactor shutdown was calculated. On the conversion, limit of fuel burn-up was increased from 50% to 60%. And the fuel exchange procedure was changed from the six-batch dispersion procedure to the fuel burn-up management procedure. The previous estimation of fuel burn-up was required for the planning of fuel exchange, so that the estimation was carried out by means of past operation data. Finally, a new fuel exchange procedure was proposed for effective use of fuel elements. On the procedure, burn-up of spent fuel was defined for each loading position. The average length of fuel's staying in the core can be increased by two percent on the procedure. (author)

Report on progress of researches by common utilization of JAERI nuclear facilities, for fiscal 1982

International Nuclear Information System (INIS)

1983-01-01

The utilization of the facilities in the Japan Atomic Energy Research Institute in common in 1982 has finished in active state, and the results of the researches have reached the stage of publication. The subjects of the researches spread over wide fields, and in 1982 also, extremely diversified researches were carried out. In this report, theses results were collected in one book, and it is desirable to utilize it actively. The number of the research themes is 131. In the field of general researches, the researches on radiochemistry, the utilization of radiation and the effects of irradiation were mostly carried out, while in cooperative researches, the researches were mainly concerned with nuclear reactor engineering and nuclear reactor materials. The total number of visitors was 3025. The facilities offered to the common utilization were JRR-2, JRR-3, JRR-4, Co-60 irradiation facility and others. The abstracts of the papers are reported. (J.P.N.)

Development of a small-aperture slit system for a high collimator ratio at the thermal neutron radiography facility in JRR-3

International Nuclear Information System (INIS)

Yasuda, Ryo; Nojima, Takehiro; Iikura, Hiroshi; Sakai, Takuro; Matsubayashi, Masahito

2011-01-01

A small-aperture slit system has been developed and installed to enhance the collimator ratio ('L/D') of the thermal neutron radiography facility (TNRF) in JRR-3. The degree of unsharpness on neutron images is reduced by increasing the L/D. The small-aperture slit system increased the L/D by creating a small aperture size ('D'). Image sharpness improved when the aperture size was reduced to below 10 mm by 10 mm in the TNRF. On the other hand, there was almost no difference in unsharpness on images obtained above 10 mm by 10 mm in aperture size. These results indicate that an aperture size of less than 10 mm by 10 mm should be used for high-spatial-resolution imaging at the TNRF. The beam area of the small-aperture slits was relatively small in comparison with that of a conventional large collimator, though gradually increasing with increasing aperture size. Even with an aperture size of 5 mm by 5 mm, the practical beam area for imaging examinations corresponded to around 25 mm by 20 mm, which is enough area to carry out high-spatial-resolution imaging. (author)

Introduction of sample environment equipment for neutron scattering experiments

International Nuclear Information System (INIS)

Shimojo, Yutaka; Ihata, Yoshiaki; Kaneko, Koji; Takeda, Masayasu

2013-02-01

Neutron scattering experiments have been frequently performed under variety of sample conditions, such as various temperatures, pressures, magnetic fields and stresses, and those complex conditions to fully utilize superior properties of neutron. To this aim, a number of sample environment equipment, refrigerators, furnaces, pressure cells, superconducting magnets are equipped in JRR-3 to be used for experiments. In this document, all available sample environment equipment in both JRR-3 reactor and guide halls are summarized. We hope this document would help neutron scattering users to perform effective and excellent experiments. (author)

Boron neutron capture therapy (BNCT). Recent aspect, a change from thermal neutron to epithermal neutron beam and a new protocol

International Nuclear Information System (INIS)

Nakagawa, Yoshinobu

1999-01-01

Since 1968, One-hundred seventy three patients with glioblastoma (n=81), anaplastic astrocytoma (n=44), low grade astrocytoma (n=16) or other types of tumor (n=32) were treated by boron-neutron capture therapy (BNCT) using a combination of thermal neutron and BSH in 5 reactors (HTR n=13, JRR-3 n=1, MuITR n=98, KUR n=28, JRR-2 n=33). Out of 101 patients with glioma treated by BNCT under the recent protocol, 33 (10 glioblastoma, 14 anaplastic astrocytoma, 9 low grade astrocytoma) patients lived or have lived longer than 3 years. Nine of these 33 lived or have lived longer than 10 years. According to the retrospective analysis, the important factors related to the clinical results were tumor dose radiation dose and maximum radiation dose in thermal brain cortex. The result was not satisfied as it was expected. Then, we decided to introduce mixed beams which contain thermal neutron and epithermal neutron beams. KUR was reconstructed in 1996 and developed to be available to use mixed beams. Following the shutdown of the JRR-2, JRR-4 was renewed for medical use in 1998. Both reactors have capacity to yield thermal neutron beam, epithermal neutron beam and mixed beams. The development of the neutron source lead us to make a new protocol. (author)

Activity report on neutron scattering research

International Nuclear Information System (INIS)

Nagao, M.; Tawata, N.; Fujii, Y.

1998-01-01

The experiments performed on the thirteen university-owned spectrometers installed at JRR-3M of JAERI in the fiscal year of 1997 were described in this report. The latest ''Neutron News'' (vol. 9, issue 3, 1998) has featured highlights of the activities based on the JRR-3M and its cover displays a graph showing an endless increase of the number of proposals to the users program in the fiscal 1997. The university-owned spectrometers are available for general users all over Japan. The users' requirement for a higher flux beam reactor became larger and larger with time. Thus, JAERI has refurbished JRR-3 to satisfy these demands. In 1997, a joint project between Chiba University and Institute for Solid State Physics (ISSP) started to build a new 4-cycle diffractometer for crystal physics/chemistry at T 2-2 beam port on a thermal guide. (M.N.)

Progress report on neutron scattering research. April 1, 2002 - March 31, 2003

International Nuclear Information System (INIS)

Koizumi, Satoshi; Matsuda, Masaaki; Kurihara, Kazuo

2004-03-01

This issue summarizes research progress in neutron scattering at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor JRR-3 during the period between April 1, 2002 and March 31, 2003. (author)

Development of radioisotope production in JAERI

International Nuclear Information System (INIS)

Yamabayashi, H.; Kato, H.; Umezawa, H.

1992-01-01

Since 1962, we have been developing methods and technology for producing a wide variety of processed radioisotopes and sealed radiation sources by using the JAERI's reactors, JRR-2, JRR-3, JRR-4 and JMTR, and providing the products to domestic users. At present, 29 nuclides and 31 products are on our list of processed radioisotopes. Some of those isotopes such as P-32, S-35, Cr-51 and short-lived nuclides are being produced regularly for distribution, but most of the rest are produced upon request. The radiation sources of Co-60 needles and Ir-192 pellets for industrial use and Gd-153 pellet, 7 kinds of Ir-192 and Au-198 grain for medical applications are produced and distributed routinely. (author)

Progress report on neutron scattering research. April 1, 1999 - March 31, 2000

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Satoru; Aizawa, Kazuya; Katano, Susumu (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

This issue summarizes research progress in neutron scattering at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor (JRR-3M) during the period between April 1, 1999 and March 31, 2000. (author)

Progress report on neutron scattering research. April 1, 2000 - March 31, 2001

International Nuclear Information System (INIS)

Aizawa, Kazuya; Katano, Susumu; Koizumi, Satoshi

2002-03-01

This issue summarizes research progress in neutron scattering at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor (JRR-3M) during the period between April 1, 2000 and March 31, 2001. (author)

Progress report on neutron scattering research. April 1, 2001 - March 31, 2002

Energy Technology Data Exchange (ETDEWEB)

Katano, Susumu; Koizumi, Satoshi; Matsuda, Masaaki (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-11-01

This issue summarizes research progress in neutron scattering at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor JRR-3M during the period between April 1, 2001 and March 31, 2002. (author)

Progress report on neutron scattering research (April 1, 1998 - March 31, 1999)

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Jun-ichi; Fujiwara, Satoru; Aizawa, Kazuya [eds.] [Advanced Science Research Center (Tokai Site), Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2000-03-01

The present issue summarizes research progress in neutron scattering at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor (JRR-3M) during the period between April 1, 1998 and March 31, 1999. (author)

Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

On Dwarves and Scientists: Probing for technological ethics in the creative imagination of J.R.R. Tolkien

OpenAIRE

Kidwell, Jeremy

2009-01-01

In this paper, I look more closely at J.R.R. Tolkien's writing to probe for a broader philosophy of technology. I suggest that (after a survey of his letters and literature), according to Tolkien none of the activities of technology, inquiry, making or use can be morally neutral. As a result, in Tolkien's account careful scrutiny must be made to ensure that these three types of technological activity occur within an ethically ordered economy, directed by appreciation for the intrinsic good of...

Construction of elements data base of JRR-3, 4 reactor

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Akira; Iwata, Yoshihiro [Akita Univ. (Japan); Sawahata, Hiroyuki

1998-01-01

Two kinds of the basic data base were constructed. One of them contains stable isotope information (atomic number, mass number, atomic weight and isotope ratio) and other consists of radionuclide information (atomic number, mass number, half-life, {gamma}-ray average energy, {beta}-ray average energy and {beta}+ray release rate). We tried to add the spectrum data determined by experiments, but many kinds of radionuclide in the sample could not separated. (S.Y.)

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

Annual report of the Japan Atomic Energy Research Institute for fiscal 1992

International Nuclear Information System (INIS)

1993-01-01

Japan Atomic Energy Research Institute has promoted the research on high temperature engineering, the research and development of nuclear fusion, the research on radiation utilization and the research and development of nuclear powered ships as the advanced project researches which bring about the breakthrough of atomic energy technology as well as the research on the safety, following the long term plan of atomic energy development and utilization which was decided in 1987, as the general research institute in Japanese atomic energy field. The progress of the above mentioned researches in fiscal 1992 is reported. The operation of JRR-2, JRR-3M, JRR-4 and JMTR was carried out as scheduled. 9 cases of the medical irradiation on brain tumors were performed at JRR-2. As to the practical test of the disassembling of JPDR, the machinery and equipment in the reactor containment vessel were removed, and the development of a high performance decontamination testing device and others was advanced. The efficient operation of the large computer system, the production and sales of radioisotopes and radioactive waste business were continued. (K.I.)

The effect of uncertainty of reactor parameters obtained using k0-NAA on result of analysis

International Nuclear Information System (INIS)

Sasajima, Fumio

2006-01-01

Neutron Activation Analysis using the k 0 method is a useful method allowing convenient and accurate simultaneous analysis of plural elements, eliminating the need for the use of comparative reference samples. As already well known, it is essential for the correct result of an analysis to obtain the α-factor and f-factor for a neutron spectrum in an irradiation field accurately when an attempt is made to use the k 0 method. For this reason, based on data obtained from the experiment conducted in the JRR-3 PN-3 system, how uncertainty of the measured values for α-factor and f-factor affects the result of an analysis was evaluated. The process of evaluation involved intentionally varying the values for reactor parameters followed by making an analysis of environmental reference samples (NIST SRM-1632c) using the k 0 method to examine the effect of these factors on the concentrations of 19 elements. The result of the evaluation revealed that the degree of the effect of uncertainty on the concentrations of 19 elements was at best approx. 1% under the condition of this experiment assuming that the factor α, a reactor parameter, had uncertainty of approx. 200%. (author)

Basic study on neutron activation analysis measuring short-lived nuclides (half-lives 0.7 to 100 s) using JRR-3M NAA facility

International Nuclear Information System (INIS)

Yonezawa, Chushiro; Ichimura, Shigeju; Matsue, Hideaki; Kurosawa, Tatsuya

1998-11-01

Analytical basis for neutron activation analysis (NAA) measuring nuclides of second order half-life produced by (n, γ) reaction has been studied using a neutron activation analysis facility of JRR-3M. Basic experimental conditions such as high count rate gamma-ray measurement, effects of irradiation capsule material and stability of neutron flux were examined. The analytical sensitivities and detection limits for 20 elements of which activated radionuclides have half-lives from 0.7 to 100 s were obtained. Scandium, Hf, Dy and In were elements having the highest analytical sensitivity, with detection limits down to 4.2 to 14 ng. Fluorine, of which determination by other methods is difficult, can be detected in more than 530 ng. Determination of ppm levels of F in silicon nitride powder using a single and cyclic activation methods were performed. Accuracy and precision for F determination were verified by analyzing reference materials of Opal Glass (NIST SRM91) and Oyster Tissue (NIST SRM1566a). The relationship between the detection limit of F and Al contents was also clarified. Analytical applications of high sensitive elements such as Se, Sc, Hf, In and Dy in various materials, including reference materials, were also examined and the accuracy, precision and detection limits of the present method were evaluated. (author)

Progress report on neutron scattering research (April 1, 1997 - March 31, 1998)

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Jun-ichi; Fujiwara, Satoru; Aizawa, Kazuya [eds.

1999-02-01

The present issue summarizes research progress in neutron scattering at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor (JRR-3M) during the period between April 1, 1997 and March 31, 1998. The 76 papers are indexed individually. (J.P.N.)

Dynamics of TRIGA-3 Salazar Reactor

International Nuclear Information System (INIS)

Gallardo S, L.F.

1990-01-01

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

COOLOD-N: a computer code, for the analyses of steady-state thermal-hydraulics in plate-type research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1990-02-01

The COOLOD-N code provides a capability for the analysis of the steady-state thermal-hydraulics of research reactors in which plate-type fuel is employed. This code is revised version of the COOLOD code, and is applicable not only to a forced convection cooling mode, but also to a natural convection cooling mode. In the code, a function to calculate flow rate under a natural convection, and a heat transfer package which was a subroutine program to calculate heat transfer coefficient, ONB temperature and DNB heat flux, and was especially developed for the upgraded JRR-3, have been newly added to the COOLOD code. The COOLOD-N code also has a capability of calculating the heat flux at onset of flow instability as well as DNB heat flux. (author)

Report of results of joint research using facilities in Japan Atomic Energy Research Institute in fiscal year 1987

International Nuclear Information System (INIS)

1988-06-01

The total themes of the joint research in fiscal year 1987 were 127. These are shown being classified into the general joint research in Tokai and Takasaki, neutron diffraction research and cooperative research. The general joint research is the standard utilization form using research reactors JRR-2 and JRR-4, Co-60 gamma irradiation facilities in Tokai and Takasaki, an electron beam irradiation facility in Takasaki, an electron beam linear accelator and hot laboratories, which are opened for common utilization by Japan Atomic Energy Research Institute. The cooperative research is carried out by concluding research cooperation contracts between the researchers of universities and JAERI. In the general joint research, radioactivation analysis, radiation chemistry, irradiation effect, neutron diffraction and so on are the main themes, and in the cooperative research, reactor technology, reactor materials, nuclear physics measurement and others are the main themes. The total number of visitors was 2629 man-day, and decreased due to the stop of JRR-2. Also other activities are reported. The abstracts of respective reports are collected in this book. (Kako, I.)

Progress report on neutron scattering research (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

Ishii, Yoshinobu; Suzuki, Jun-ichi; Fujiwara, Satoru [eds.

1997-10-01

This issue summarizes research progress in neutron scattering at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor (JRR-3M) mainly during the period between April 1, 1996 and March 31, 1997. The 57 of the presented papers are indexed individually. (J.P.N.)

Present status of radioisotope production in JAERI

International Nuclear Information System (INIS)

Yamabayashi, Hisamichi

1994-01-01

Since 1962, the technology for producing a wide variety of processed radioisotopes and sealed radiation sources has been developed by using the reactors, JRR-1, JRR-2, JRR-3, JRR-4 and JMTR, and the products have been offered to domestic users. At present, 31 products of 29 nuclides are on the list of processed radioisotopes. Some of those isotopes such as P-32, S-35, Cr-51 and short lived nuclides are produced for regular distribution, but the rest are produced upon request. The radiation sources of Co-60 needles for industrial use, Ir-192 pellets for the nondestructive inspection of pipelines, Gd-153 pellets for the diagnosis of born mineral and seven kinds of brachy therapy Ir-192 and Au-198 grains are produced and distributed regularly. The organic compounds labeled with H-3 and C-14 are widely used. In fiscal year 1992, 34 batches and total amount 12 TBq of processed radioisotopes and 100 batches, 1.2 PBq of radiation sources were produced as scheduled. The development of the techniques for producing the sources emitting high energy β ray used for the diagnosis and treatment of cancer is in progress. The method of producing new isotopes is developed. (K.I.)

Proceedings of the 5th workshop of 'quantum complex phenomena, -superconductivity, magnetism and phonon-' under the NIMS-RIKEN-JAEA cooperative research program on quantum beam science and technology

International Nuclear Information System (INIS)

Shamoto, Shin-ichi; Kodama, Katsuaki

2012-08-01

The 5th workshop of the NIMS-RIKEN-JAEA Cooperative Research Program 'Quantum Complex Phenomena, -Superconductivity, Magnetism and Phonon-' was held on January 23-24, 2012 at KKR Hakone Miyanoshita. This workshop is aimed to reveal the mechanism of quantum complex phenomena for the developments of next generation functional materials on the basis of 'Joint Research Agreement for the Pioneering R and D with Quantum Beam Technology' concluded by NIMS, RIKEN and JAEA on December 20, 2006. The neutron facilities in Tokai, i.e., the research reactor JRR-3 and the proton accelerator J-PARC (Japan Proton Accelerator Research Complex), were damaged by the Tohoku-Kanto earthquake on March 11, 2011. J-PARC members' devoted efforts for the recovery made it possible to successfully produce neutron beam in the midnight of January 24, just after this workshop. Also at JRR-3 all the repair works of the reactor facilities and buildings have been completed by the end of the fiscal year 2011. Yet, the safety analysis report is to be submitted and after its positive review by the national regulatory authority, the JRR-3 can undergo the regular periodic inspection to resume its operation. Under this circumstance, characteristic technologies, instruments, and distinguished researches of each institute about 'Superconductivity, Magnetism and Phonon' are introduced and discussed in addition to research outcomes of this Joint Research Agreement including a future prospect of this research area. This report includes abstracts and materials of the presentations in the workshop. (author)

Progress report on neutron science. April 1, 2004 - March 31, 2005

International Nuclear Information System (INIS)

Kurihara, Kazuo; Moriai, Atsushi; Matsuda, Masaaki

2005-09-01

This issue summarizes research progress in neutron science at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor JRR-3 during the period between April 1, 2004 and March 31, 2005. This report contains highlights in research by 13 neutron research groups at JAERI and 91 summary articles. (author)

Progress report on neutron science. April 1, 2005 - March 31, 2006

International Nuclear Information System (INIS)

Moriai, Atsushi; Takeda, Masayasu; Kurihara, Kazuo

2007-03-01

This issue summarizes research progress in neutron science at Japan Atomic Energy Agency (JAEA) by utilizing the research reactor JRR-3 during the period between April 1, 2005 and March 31, 2006. This report contains highlights in research by 13 neutron research groups at JAEA and 89 summary articles. (author)

The development of the measurement technique of the control rod worth with the inverse kinetics method considering the influence of the steady neutron source

International Nuclear Information System (INIS)

Takeuchi, Mitsuo; Wada, Shigeru; Takahashi, Hiroyuki; Hayashi, Kazuhiko; Murayama, Yoji

2000-09-01

At the research reactor such as JRR-3M, the operation management is carried out in order to ensure safe operation, for example, the excess reactivity is measured regularly and confirmed that it satisfies a safety condition. The excess reactivity is calculated using control rod position in criticality and control rod worth measured by a positive period method (P.P method), the conventional inverse kinetic method (IK method) and so on. The neutron source, however, influences measurement results and brings in a measurement error. A new IK method considering the influence of the steady neutron sources is proposed and applied to the JRR-3M. This report shows that the proposed IK method measures control rod worth more precisely than a conventional IK method. (author)

Experimental study of neutron streaming through steel-walled annular ducts in reactor shields

International Nuclear Information System (INIS)

Toshimas, M.; Nobuo, S.

1983-01-01

For the purpose of providing experimental data to assess neutron streaming calculations, neutron flux measurements were performed along the axes of the steel-walled annular ducts set up in a water shield of the pool-type reactor JRR-4. An annular duct simulated the air gap around the main coolant pipe. Another duct simulated the streaming path around the primary circulating pump of the integrated-type marine reactor. A 90-deg bend annular duct was also studied. In a set of measurements, the distance Z between the core center and the duct axis and the annular gap width delta were taken as parameters, that is, Z = 0, 80, and 160 cm and delta = 2.2, 4.7, and 10.1 cm. The reaction rates and the fluxes measured by the activation method are given in terms of absolute magnitude within an accuracy of + or - 30%. An empirical formula is derived based on those measured data, which describes the axial distribution of the neutron flux in the steel-walled annular duct in reactor shields. It is expressed by a simple function of the axial distance in units of the square root of the line-of-sight area, S /SUB l/ . The accuracy of the formula is examined by taking into account the duct location with respect to the reactor core, the neutron energy, the steel wall thickness, and the media outside of the steel wall. The accuracy of the formula is, in general, <30% in the axial distance between 3√S /SUB l/ and 30√S /SUB l/

Progress report on neutron science. April 1, 2003 - March 31, 2004

International Nuclear Information System (INIS)

Matsuda, Masaaki; Kurihara, Kazuo; Moriai, Atsushi

2005-03-01

This issue summarizes research progress in neutron science at Japan Atomic Energy Research Institute (JAERI) by utilizing the research reactor JRR-3 during the period between April 1, 2003 and March 31, 2004. This report contains highlights in research from 10 neutron research groups at JAERI and summary reports of 82 papers. (author)

Establishment of the observing system for boron in steels by alpha-particle track etching method using JAERI reactor

International Nuclear Information System (INIS)

Asakura, Kentaro; Shibata, Koji; Sawahata, Hiroyuki; Kawate, Minoru; Harasawa, Susumu

2003-01-01

Alpha-particle track etching (ATE) method is most effective in observing boron distribution in steels. Previously, in Japan, neutron irradiation for this method was carried out in the reactor at the Institute of Atomic Energy, Rikkyo University. This reactor, however, was shut down in 1999. Therefore, the establishment of a new system for ATE method has been required and experimental research was performed using the reactor at the Japan Atomic Energy Research Institute (JAERI). It was clarified that the irradiation equipment for medical treatment of the reactor JRR-4 was most suitable for ATE method. The specimen trestle for low radioactive exposure was newly-developed. ATE image obtained by 12h irradiation using this trestle showed a good quality similar to that obtained using Rikkyo's reactor and that obtained using the trestle of the old model. Using this new trestle, the amount of neutron which the worker suffers during the operation at the irradiation equipment decreases from 4μSv/h to 0-1 μSv/h compared with the trestle of the old model. The total amount of thermal neutron after 12 h irradiation was almost same as that under the recommended condition of the reactor at Rikkyo University, 6.5 x 10 14 n cm -2 . (author)

MYTHONYMICAL SPACE PECULIARITIES IN FOREIGH FANTASY LITERATURE (BASED ON LITERARY WORKS BY J.R.R. TOLKIEN AND G.R.R. MARTIN

Directory of Open Access Journals (Sweden)

Kharitonova, E.V.

2017-12-01

Full Text Available The paper aims at exemplifying the characteristic features of onomastic space of a fantasy text. The authors examine the notion of a ‘mythonym’, systemize different approaches to classifying mythonymical units. The conducted study permitted to work out a thematic classification of mythonyms from the novels “The Lord of the Rings: The Fellowship of the Ring” by J.R.R. Tolkien and “A Song of Ice and Fire: A Game of Thrones” by G.R.R. Martin and to draw the conclusion on mythonyms’ role in fantasy literature.

Decommissioning of a small reactor (BR3 reactor, Belgium)

International Nuclear Information System (INIS)

Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

2002-01-01

Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

Report on the progress of researches using JAERI facilities in common, fiscal 1979

International Nuclear Information System (INIS)

1980-01-01

The utilization of the facilities in the Japan Atomic Energy Research Institute in common in 1979 has finished in active state, and the results of the researches have reached the stage of publication. The subjects of the researches spread over wide fields, and in 1979 also, extremely diversified researches were carried out. In this report, these results were collected in one book, and it is desirable to utilize it actively. It is expected that the research activities using the JAERI facilities in common will be promoted more and more widely and powerfully, but there are many problems in the manpower, equipment, space and so on required for maintaining and promoting such activities, and it is necessary to improve and strengthen the environment of researches. The number of the research themes is 125. In the field of general researches, the researches on radio-chemistry, the utilization of radiation and the effects of irradiation were mostly carried out, while in cooperative researches, the researches were mainly concerned with nuclear reactor engineering and nuclear reactor materials. The total number of visitors was 3863. The facilities offered to the common utilization were JRR-2, JRR-3, JRR-4, Co-60 irradiation facility, hot laboratory, linear accelerator, No. 1 and No. 2 electron accelerators. The abstracts of the papers are reported. (Kako, I.)

Experimental study on reactivity measurement in thermal reactor by polarity correlation method

International Nuclear Information System (INIS)

Yasuda, Hideshi

1977-11-01

Experimental study on the polarity correlation method for measuring the reactivity of a thermal reactor, especially the one possessing long prompt neutron lifetime such as graphite on heavy water moderated core, is reported. The techniques of reactor kinetics experiment are briefly reviewed, which are classified in two groups, one characterized by artificial disturbance to a reactor and the other by natural fluctuation inherent in a reactor. The fluctuation phenomena of neutron count rate are explained using F. de Hoffman's stochastic method, and correlation functions for the neutron count rate fluctuation are shown. The experimental results by polarity correlation method applied to the β/l measurements in both graphite-moderated SHE core and light water-moderated JMTRC and JRR-4 cores, and also to the measurement of SHE shut down reactivity margin are presented. The measured values were in good agreement with those by a pulsed neutron method in the reactivity range from critical to -12 dollars. The conditional polarity correlation experiments in SHE at -20 cent and -100 cent are demonstrated. The prompt neutron decay constants agreed with those obtained by the polarity correlation experiments. The results of experiments measuring large negative reactivity of -52 dollars of SHE by pulsed neutron, rod drop and source multiplication methods are given. Also it is concluded that the polarity and conditional polarity correlation methods are sufficiently applicable to noise analysis of a low power thermal reactor with long prompt neutron lifetime. (Nakai, Y.)

Neutron scattering equipments in JAERI. Current status

International Nuclear Information System (INIS)

Hamaguchi, Yoshikazu; Minakawa, Nobuaki

2003-01-01

24 neutron scattering instruments are installed in the JRR-3M research reactor. Among them JAERI has 12 neutron scattering instruments. Those instruments are HRPD for high-resolution structural analysis, TAS-1 and TAS-2 for elastic and inelastic scattering and for magnetic scattering measurements by the polarized neutron, LTAS for elastic and inelastic scattering measurement at a low energy region, and for neutron device development, PNO for topography and for very small angle scattering measurement in a small Q range, NRG for neutron radiography, RESA for internal strain measurements, SANS for the molecule and semi-macroscopic magnetic structural analysis, BIX-2 and BIX-3 for the biological structural analysis research, and PGA for the research of prompt gamma-ray analysis. The university groups have 12 neutron scattering instruments. Since those instruments were installed at the period when JRR-3M was completed, about 10 years have passed. In order to match the old control systems with the progress of recent computer technologies, and peripheral equipment, numbers of instruments are being renewed. In the neutron guide hall of JRR-3M, the Ni mirror guide tube was replaced by a super mirror guide tube to increase neutron flux. The intensity of 2A flux was increased by a factor of about two. (J.P.N.)

Integral test of JENDL-3.3 for fast reactors

International Nuclear Information System (INIS)

Chiba, Gou

2003-01-01

An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

Uses of reactor neutrons for studying the microcomposition of materials

International Nuclear Information System (INIS)

Jervis, R.E.

1993-01-01

Reactor neutrons constitute excellents 'probes' for exploring and measuring a wide range both of minor and trace constituents in solids and liquids with high sensitivity because of their transparency in materials. Nondestructive neutron prompt-gamma analysis (PGA) utilizing either cold or thermal neutrons, such as at JRR-3M, is compared and contrasted to the more common (delayed) instrumental neutron activation analysis (INAA) and epithermal NAA. Clearly PGA offers high sensitivity for selected elements: B, H, Cd and REE's in suitable matrices, and is therefore, complementary to INAA which is not as useful for them, or for Ni, Sn, Fe, C or N. Recent INAA applications in our laboratory that demonstrate some of the uniqueness of neutron methods include use of epithermal neutrons for small biological specimens to measure Cd, K, As, Zn and, multielemental INAA for environmental pollution studies. The latter involves large data sets of multielemental concentrations which are subjected to statistical multivariant factor analysis to reveal unknown or unsuspected quantitative relationships among groups of trace constituents. These patterns, or 'factors' are shown to be uniquely related to pollution sources and can be utilized to compute the relative source contributions at a given receptor site. (author)

Nuclear calculation for employing medium enrichment in reactors of Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Miyasaka, Yasuhiko

1979-01-01

The fuel used for the research reactors of Japan Atomic Energy Research Institute (JAERI) is presently highly enriched uranium of 93%. However, the U.S. government (the supplier of fuel) is claiming to utilize low or medium enriched uranium from the viewpoint of resistivity to nuclear proliferation, and the availability of highly enriched uranium is becoming hard owing to the required procedure. This report is described on the results of nuclear calculation which is the basis of fuel design in the countermeasures to the reduction of enrichment. The basic conception in the reduction of enrichment is three-fold: to lower the latent potential of nuclear proliferation as far as possible, to hold the present reactor performance as far as possible, and to limit the reduction in the range which is not accompanied by the modification of reactor core construction and cooling system. This time, the increase of the density and thickness of fuel plates and the effect of enrichment change to 45% on reactivity and neutron flux were investigated. The fuel of UAl sub(x) - Al system was assumed, which was produced by powder metallurgical method. The results of investigations on JRR-2 and JMTR reactors revealed that 45% enriched fuel does not affect the performances much. However, deterioration of the performances is not neglegible if further reduction is needed. In future, the influence of the burn-up effect of fuel on the life of reactor cores must be investigated. (Wakatsuki, Y.)

Crystal structures of new cuprates containing CO3 analyzed by the Rietveld method of neutron powder diffraction

International Nuclear Information System (INIS)

Miyazaki, Y.; Yamane, H.; Kajitani, T.; Hiraga, K.; Hirai, T.; Morii, Y.; Funahashi, S.

1993-01-01

New compounds containing CO 3 groups, Sr 2 CuO 2 (CO 3 ), (C 0.4 Cu 0.6 )Sr 2 (Y 0.86 Sr 0.14 )Cu 2 O 7 and (C 0.35 Cu 0.65 )Sr 2 (Y 0.73 Ce 0.27 ) 2 Cu 2 O 9 , were prepared as stable phases at 1273-1303 K in a flowing gas of O 2 -CO 2 . The crystal structures of these compounds were refined by means of the Rietveld analysis for neutron powder diffraction data collected using a high resolution powder diffractometer (HRPD) in the JRR-3M reactor hall of the Japan Atomic Energy Research Institute (JAERI). Positions of CO 3 groups were satisfactorily determined. The distances of C-O bonds in the CO 3 groups were around 1.3A and the O-C-O angles were almost equal to the ideal bond angle of 120deg. (author)

Masculine and Feminine Insights Into the Fantastic World of Elves: J.R.R. Tolkien’s the Lord of the Rings and Muriel Barbery’s the Life of Elves

Directory of Open Access Journals (Sweden)

Măcineanu Laura

2016-12-01

Full Text Available The paper focuses on the fantasy universes created by J.R.R. Tolkien and by the more recent French writer, Muriel Barbery. The two authors excel in their depiction of the elusive world of the elves, each offering a deeply personal vision (a man writer’s and a woman writer’s of what such magical beings may be like and how they may relate to humankind.

Measurements of neutron cross section of the {sup 243}Am(n,{gamma}){sup 244}Am reaction

Energy Technology Data Exchange (ETDEWEB)

Hatsukawa, Yuichi; Shinohara, Nobuo; Hata, Kentaro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

The effective thermal neutron cross section of {sup 243}Am(n,{gamma}){sup 244}Am reaction was measured by the activation method. Highly-purified {sup 243}Am target was irradiated in an aluminum capsule by using a research reactor JRR-3M. The tentative effective thermal neutron cross sections are 3.92 b, and 84.44 b for the production of {sup 244g}Am and {sup 244m}Am, respectively. (author)

Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

Energy Technology Data Exchange (ETDEWEB)

Gallardo S, L F

1991-12-31

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

Energy Technology Data Exchange (ETDEWEB)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

2016-09-15

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Application of nuclear pumped laser to an optical self-powered neutron detector

Science.gov (United States)

Yamanaka, N.; Takahashi, H.; Iguchi, T.; Nakazawa, M.; Kakuta, T.; Yamagishi, H.; Katagiri, M.

1996-05-01

A Nuclear Pumped Laser (NPL) using 3He/Ne/Ar gas mixture is investigated for a purpose of applying to an optical self-powered neutron detector. Reactor experiments and simulations on lasing mechanism have been made to estimate the best gas pressure and mixture ratios on the threshold input power density (or thermal neutron flux) in 3He/Ne/Ar mixture. Calculational results show that the best mixture pressure is 3He/Ne/Ar=2280/60/100 Torr and thermal neutron flux threshold 5×1012 n/cm2 sec, while the reactor experiments made in the research reactor ``YAYOI'' of the University of Tokyo and ``JRR-4'' of JAERI also demonstrate that excitational efficiency is maximized in a similar gas mixture predicted by the calculation.

Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

International Nuclear Information System (INIS)

Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

2013-01-01

The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

Kernbichler, W.; Miley, G.H.; Heindler, M.

1989-01-01

The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

Development of the external cooling device of increase the productivity of neutron-transmutation-doped silicon semiconductor (NTD-Si) (Joint research)

International Nuclear Information System (INIS)

Hirose, Akira; Wada, Shigeru; Sasajima, Fumio; Kusunoki, Tsuyoshi; Kameyama, Iwao; Aizawa, Ryouji; Kikuchi, Naoyuki

2007-01-01

Neutron-Transmutation-Doped Silicon Semiconductor (hereinafter referred as 'NTD-Si') is the best semiconductor for the power device. The needs of NTD-Si increase recently in proportion to the popularization of hybrid-cars. A fission research reactor, which is a steady state neutron source, is being expected as the best device to meet the needs. So far, we have reconsidered the existing approach which is employed for NTD-Si production works at the research reactors JRR-3, JRR-4 and JMTR of JAEA so as to meet the needs. As one of the effective measures, we found out that the productivity can be increased by incorporating a new device to cool down radioactivity of irradiated silicon ingots at the place outside the main stream from the loading of silicon ingots to the withdrawal of irradiated ingots to the existing JRR-3 Uniformity Irradiation System. Consequently, we developed and installed the device (hereinafter referred as 'external cooling device'). After an ingot was irradiated once, it is turned over manually and irradiated again in order to irradiate the ingot uniformly. With the conventional system, it was necessary to wait the radioactivity of ingot decrease less than the permissible level with holding the ingot in the irradiation equipment. It was effective to shorten the waiting period by using an external cooling device for production increase of NTD-Si. It is expected that the productivity of NTD-Si will be increased by using the external cooling device. This report mentions the design of the external cooling device and verification between its design specifications and the performance of the device completed. (author)

Measurements of neutron cross sections of radioactive waste nuclides

Energy Technology Data Exchange (ETDEWEB)

Katoh, Toshio [Gifu College of Medical Technology, Seki, Gifu (Japan); Harada, Hideo; Nakamura, Shoji; Tanase, Masakazu; Hatsukawa, Yuichi

1998-01-01

Accurate nuclear reaction cross sections of radioactive fission products and transuranic elements are required for research on nuclear transmutation methods in nuclear waste management. Important fission products in the nuclear waste management are {sup 137}Cs, {sup 135}Cs, {sup 90}Sr, {sup 99}Tc and {sup 129}I because of their large fission yields and long half-lives. The present authors have measured the neutron capture cross sections and resonance integrals of {sup 137}Cs, {sup 90}Sr and {sup 99}Tc. The purpose of this study is to measure the neutron capture cross sections and resonance integrals of nuclides, {sup 129}I and {sup 135}Cs accurately. Preliminary experiments were performed by using Rikkyo University Reactor and JRR-3 reactor at Japan Atomic Energy Research Institute (JAERI). Then, it was decided to measure the cross section and resonance integral of {sup 135}Cs by using the JRR-3 Reactor because this measurement required a high flux reactor. On the other hand, those of {sup 129}I were measured at the Rikkyo Reactor because the product nuclides, {sup 130}I and {sup 130m}I, have short half-lives and this reactor is suitable for the study of short lived nuclide. In this report, the measurements of the cross section and resonance integral of {sup 135}Cs are described. To obtain reliable values of the cross section and resonance integral of {sup 135}Cs(n, {gamma}){sup 136}Cs reaction, a quadrupole mass spectrometer was used for the mass analysis of nuclide in the sample. A progress report on the cross section of {sup 134}Cs, a neighbour of {sup 135}Cs, is included in this report. A report on {sup 129}I will be presented in the Report on the Joint-Use of Rikkyo University Reactor. (author)

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

Determination of α and f for k0-NAA in irradiation sites with high thermalized neutrons

International Nuclear Information System (INIS)

Ho Manh Dung; Fumio Sasajima

2003-01-01

At the irradiation sites in a research reactor where neutrons are well thermalized, the 'bare' irradiation methods for α and f determinations that are incorporated in the KAYZERO R /SOLCOI R v4.0 package seem to be unworkable. It is a fact that when carrying out the determination of α and f at the PN-3 irradiation site of the JRR-3M research reactor (Tokai, Japan) using the KAYZERO software, unacceptable results were obtained from experiment to experiment for α ranging from 0.10 to 0.999 and f ranging from 70 to 9999. The 'Cd ratio' for a gold monitor in the PN-3 site is about 300. These unacceptable results have indicated the need for the application of Cd methods that are based on 'Cd-covered' irradiation and 'Cd-ratio' measurements at certain suitable and available monitors. Such Cd methods have not been incorporated into the KAYZERO v4.0 software. Therefore, computational and experimental procedures of the 'Cd-covered' method for α and the 'Cd-ratio' method for f determination have been presented in this paper. The PN-2 irradiation site of the JRR-3M reactor where the neutron thermalization is not high has a Cd ratio for gold of about 26. The commonly used 'bare' irradiation methods can be applied together with the Cd methods so that all results may be presented as comparable data. The results of α and f determinations that were carried out in the PN-2 and PN-3 irradiation sites are presented. (author)

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

JAERI RERTR program

International Nuclear Information System (INIS)

Sato, M.

1983-01-01

In 1979, Japan Atomic Energy Research Institute (JAERI) has started a five year program for the utilization of reduced enrichment uranium fuel, in place of currently used highly enriched uranium fuel for the JAERI Research and Test Reactors (RERTR), such as the JRR-2 (10 MWt), the JRR-4 35 MWt), the JMTR(50 MWO and the JMTRC (100 Wt, nuclear mockup of the JMTR), for contributing to the reduction of proliferation concerns. In order to assess the feasibility of converting the JAERI reactors to use of fuel with reduced enrichment uranium, ANL and JAERI have embarked on a joint study program since January 1980. This document provides outlines of the JAERI RERTR Program and of the ANL-JAERI Joint Study, and main results of JAERI's Phase A report which was made under the joint study

Present status of irradiation tests on tritium breeding blanket for fusion reactor

International Nuclear Information System (INIS)

Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

1994-01-01

To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

EL3 reactor description and safety analysis report

International Nuclear Information System (INIS)

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10 14 neutrons/cm 2 /sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements [fr

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2017-07-15

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

Development of an automatic prompt gamma-ray activation analysis system

International Nuclear Information System (INIS)

Osawa, Takahito

2013-01-01

An automatic prompt gamma-ray activation analysis system was developed and installed at the Japan Research Reactor No. 3 Modified (JRR-3M). The main control software, referred to as AutoPGA, was developed using LabVIEW 2011 and the hand-made program can control all functions of the analytical system. The core of the new system is an automatic sample exchanger and measurement system with several additional automatic control functions integrated into the system. Up to fourteen samples can be automatically measured by the system. (author)

Instruments and accessories for neutron scattering research

International Nuclear Information System (INIS)

Ishii, Yoshinobu; Morii, Yukio

2000-04-01

This report describes neutron scattering instruments and accessories installed by four neutron scattering research groups at the ASRC (Advanced Science Research Center) of the JAERI and the recent topics of neutron scattering research using these instruments. The specifications of nine instruments (HRPD, BIX-I, TAS-1 and PNO in the reactor hall, RESA, BIX-II, TAS-2, LTAS and SANS-J in the guide hall of the JRR-3M) are summarized in this booklet. (author)

Development of Neutron Interferometer with Wide-Gapped ''BSE''s for Precision Measurements

International Nuclear Information System (INIS)

Seki, Y.; Kitaguchi, M.; Hino, M.; Funahashi, H.; Taketani, K.; Otake, Y.; Shimizu, H. M.

2007-01-01

We are developing large-dimensional cold-neutron interferometers with multilayer mirrors in order to investigate small interactions. In particular Jamin type interferometers composed of wide-gapped 'BSE's, which divide the beam completely, can realize the precision measurement of topological Aharonov-Casher effect. We have made a prototype with 200 μm gapped BSEs and confirmed the spatial separation of its two paths at monochromatic cold-neutron beamline MINE2 on JRR-3M reactor in JAEA

Practical use of control rod calibration system with the inverse kinetics method

International Nuclear Information System (INIS)

Yamanaka, Haruhiko; Hayashi, Kazuhiko; Motohashi, Jun; Kawashima, Kazuhito; Ichimura, Toshiyuki; Tamai, Kazuo; Takeuti, Mitsuo

2002-01-01

The control rod calibration results in the JRR-3 are used as a reactivity standard to measure and manage the reactivity change in the core. The total travel of all six control rods has been calibrated by an inverse kinetics method (IK method) during an annual maintenance period. The IK method has the great merit in saving measuring time compared with the conventional positive period method (PP method). The JRR-3 control rod calibration system was renovated and put into practical use in order to improve reliability and function by accumulating 10-year experience with the IK method in the JRR-3. The report shows the function, the performance and results of verification of the JRR-3 control rod calibration system. (author)

3 Investment Scenarios for Fast Reactors

International Nuclear Information System (INIS)

Shoai Tehrani, Bianka; Da Costa, Pascal

2013-01-01

Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

D-3He fueled FRC reactor 'ARTEMIS-L'

International Nuclear Information System (INIS)

Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

1992-09-01

A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

Description of the RA-3 research reactor as a model facility

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

Human resource development in the beginning phase of nuclear technology development in Japan

International Nuclear Information System (INIS)

Yamashita, Kiyonobu

2015-01-01

Japan Research Reactor No.1 (JRR-1) was constructed as the first nuclear reactor in Japan and reached the first criticality in 1957. The construction of both the first BWR and the first PWR were started in the same year 1967 and they started power operation in the same year 1970. Engineers of electrical utilities and vendors gave efforts to have knowledge for reactor engineering mainly on the job training with high self-motivation to contribute for nuclear technology development. A part of them participated in the reactor engineering training course of the JAERI. (author)

The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

1999-01-01

Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

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)

3D CAD model of the subcritical nuclear reactor of IPN

International Nuclear Information System (INIS)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

2016-09-01

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Modelling of MOCVD Reactor: New 3D Approach

Science.gov (United States)

Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

2014-04-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Modelling of MOCVD reactor: new 3D approach

International Nuclear Information System (INIS)

Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

2014-01-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

Science.gov (United States)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

Operating reactors licensing actions summary. Vol. 3, No. 3

International Nuclear Information System (INIS)

1983-04-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

Report on generation IV technical working group 3 : liquid metal reactors

International Nuclear Information System (INIS)

Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

2002-01-01

This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

EL-3 dismantling of an experimental reactor

International Nuclear Information System (INIS)

1989-01-01

The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

1992-01-01

The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

Characteristics of D-3He fueled frc reactor: ARTEMIS-L

International Nuclear Information System (INIS)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

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

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-06-01

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

Annual report of the Japan Atomic Energy Research Institute, Apr. 1987 - Mar. 1988

International Nuclear Information System (INIS)

1988-01-01

In fiscal 1987, the Atomic Energy Commission of Japan revised the 'Long term program for development and utilization of nuclear energy', which is done every five years. It determined the future direction of the research and development in JAERI. Three basic objectives are indicated in it, that is, to qualitatively upgrade nuclear technology, to tackle the frontiers of the technology, and to promote international collaboration. The new 'Long term task plan' of JAERI based on the 'Long term program' was established in August, 1987. The integral experiment on loss of coolant accident, the test on reactivity-initiated accident, the irradiation and post-irradiation examination of fuel, the development of safety analysis codes, the design and safety analysis of the Nuclear Fuel Cycle Safety Engineering Research Facility, the treatment and disposal of radioactive wastes, the preparatory work for constructing the High Temperature Engineering Test Reactor, the full scale plasma heating experiment with JT-60, the study on high efficiency confinement mode and current drive, the construction of the new home port for the nuclear ship 'Mutsu', the design of an advanced marine reactor, the reactor physics of FBRs and ATRs and other fundamental studies, and the operation of JRR-2, JRR-4 and JMTR are reported. (K.I.)

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

CERN Document Server

Chiba, G

2002-01-01

It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

Development of 3D CFD simulation method in nuclear reactor safety analysis

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

Energy Technology Data Exchange (ETDEWEB)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

International Nuclear Information System (INIS)

Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

1982-03-01

A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Systems analysis of the CANDU 3 Reactor

Energy Technology Data Exchange (ETDEWEB)

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Ignition access in a D-3He helical reactor

International Nuclear Information System (INIS)

Mitarai, Osamu

2003-01-01

Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

KURRI progress report 2006. April 2006 - March 2007

International Nuclear Information System (INIS)

2007-01-01

The fiscal year 2006 was the first year of outage period of the Kyoto University research Reactor (KUR) because of fuel conversion procedure from the highly enriched uranium (HEU) fuel of 93% to the low enriched uranium (LEU) of 20%. Meanwhile, the Kyoto University Critical Assembly (KUCA) have been operated for 767 hours. Kyoto University Research Reactor Institute (KURRI) accepted 3,120 man-day researchers including students in total for their experiments in accordance with the joint use programs among Japanese universities and public research organizations, and for scientific meetings held at KURRI. In addition, 104 and 139 man-day researchers have been sent to the research reactor HANARO in Korea Atomic Energy Research Institute (KAERI) and the research reactor JRR-4 in Japan Atomic Energy Agency (JAEA), respectively, in the framework of project research in the joint use program at KURRI. The results of these activities are reported in this progress report. The total 98 research issues are categorized to the 3 types of Project Research, Collaboration Research and Original Research. Project Research includes the 9 studies as follows; (1) Production of radioactive nuclei and their application to material science. (2) Fundamental study on radiation protection of fixed-field alternating-gradient accelerator. (3) Study on the effects of structure and function of proteins by irradiation and repair mechanism. (4) Study of materials irradiation effects and improvement of irradiation fields. (5) Study on neutron reaction cross sections using various neutron sources. (6) Nuclear physical and chemical characteristics of actinide and fission product nuclides. (7) Clinical study of boron neutron capture therapy as JRR-4 project. (8) Neutron activation analysis experiments using HANARO as HANARO project. (9) Development and application of new technology on neutron imaging as HANARO project. Collaboration Research includes the 8 topics as follows; (1) Slow Neutron Physics and

Fusion blankets for catalyzed D--D and D--He3 reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

Fusion blankets for catalyzed D--D and D--3He reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Support system for Neutron Activation Analysis

International Nuclear Information System (INIS)

Sasajima, Fumio; Ohtomo, Akitoshi; Sakurai, Fumio; Onizawa, Koji

1999-01-01

In the research reactor of JAERI, the Neutron Activation Analysis (NAA) has been utilized as a major part of an irradiation usage. To utilize NAA, research participants are always required to learn necessary technique. Therefore, we started to examine a support system that will enable to carry out INAA easily even by beginners. The system is composed of irradiation device, gamma-ray spectrometer and data analyzing instruments. The element concentration is calculated by using KAYZERO/SOLCOI software with the K 0 standardization method. In this paper, we review on a construction of this INAA support system in JRR-3M of JAERI. (author)

Validation of JENDL-3.3 by criticality benchmark testing

International Nuclear Information System (INIS)

Takano, Hideki; Nakagawa, Tsuneo

2001-01-01

In the thermal uranium core, the keff-values of STACY, TRACY and JRR-4 overestimated with JENDL-3.2 were improved significantly by decreasing of about 0.6% with JENDL-3.3. This is due to modification of the fission spectrum and thermal fission cross section of 235 U from JENDL-3.2 to JENDL-3.3 data. For the uranium fast cores, the discrepancies of keff values between JENDL-3.2 and 3.3 were very small. In the thermal Pu cores of TCA, the keff-values calculated with JENDL-3.3 were in good agreement with the experimental values. For Pu fuel cores of ZPPR-9 and FCA-XVII, the keff values calculated with JENDL-3.3 became larger 0.2% than those for JENDL-3.2. In small fast cores with U-233 fuel, the keff-values overestimated with JENDL-3.2 were improved considerably with JENDL-3.3, due to reevaluation of U-233 fission cross sections in the high energy region. (author)

GRIMH3: A new reactor calculation code at Savannah River Site

International Nuclear Information System (INIS)

Le, T.T.; Pevey, R.E.

1993-01-01

The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Development for advanced materials and testing techniques

Energy Technology Data Exchange (ETDEWEB)

Hishinuma, Akimichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Recent studies using a JMTR and research reactors of JRR-2 and JRR-3 are briefly summarized. Small specimen testing techniques (SSTT) required for an effective use of irradiation volume and also irradiated specimens have been developed focussing on tensile test, fatigue test, Charpy test and small punch test. By using the small specimens of 0.1 - several mm in size, similar values of tensile and fatigue properties to those by standard size specimens can be taken, although the ductile-brittle transition temperature (DBTT) depends strongly on Charpy specimen size. As for advanced material development, R and D about low activation ferritic steels have been done to investigate irradiation response. The low activation ferritic steel, so-called F82H jointly-developed by JAERI and NKK for fusion, has been confirmed to have good irradiation resistance within a limited dose and now selected as a standard material in the fusion material community. It is also found that TiAi intermetallic compounds, which never been considered for nuclear application in the past, have an excellent irradiation resistance under an irradiation condition. Such knowledge can bring about a large expectation for developing advanced nuclear materials. (author)

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

International Nuclear Information System (INIS)

Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

2012-01-01

10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

Status of decommissioning and waste management in the Nuclear Science Research Institute of JAEA

International Nuclear Information System (INIS)

Okoshi, Minoru; Yamashita, Toshiyuki

2007-01-01

The Nuclear Science Research Institute (NSRI) of JAEA has some experiences of the decommissioning of research reactors and research laboratories including a reprocessing test facility. In order to dismantle those facilities safely, we paid much attention for the radiological protection of radiation workers taking into consideration of characteristics of each facility, especially to protect internal exposures. As the results of decommissioning activities, several thousands tons of solid radioactive wastes were generated. In the near future, we will start the treatment of these stored wastes by a super compactor, metal melting furnace and non-metal waste melting furnace to gain high volume reduction and to prepare stable waste forms for final disposal. In Japan, the clearance system was established in 2005 by amending the Nuclear Regulatory Law. The NSRI plans to release very slightly contaminated concrete debris for recycling, which was generated from the replacement of reactor core of research reactor (JRR-3), according to the clearance system. (author)

Safety in the ARIES-III D-3He tokamak reactor design

International Nuclear Information System (INIS)

Herring, J.S.; Dolan, T.J.

1992-01-01

This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

Conceptual Design of Bottom-mounted Control Rod Drive Mechanism

International Nuclear Information System (INIS)

Lee, Jin Haeng; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Dongmin; Kim, Jong In

2013-01-01

The arrangement of the BMCRDMs and irradiation holes in the core is therefore easier than that of the top-mounted CRDM. Hence, many foreign research reactors, such as JRR-3M, JMTR, OPAL, and CARR, have adopted the BMCRDM concept. The purpose of this paper is to introduce the basic design concept on the BMCRDM. The major differences of the CRDMs between HANARO and KJRR are compared, and the design features and individual system of the BMCRDM for the KJRR are described. The Control Rod Drive Mechanism (CRDM) is a device to regulate the reactor power by changing the position of a Control Absorber Rod (CAR) and to shut down the reactor by fully inserting the CAR into the core within a specified time. The Bottom-Mounted CRDM (BMCRDM) for the KiJang Research Reactor (KJRR) is a quite different design concept compared to the top-mounted CRDM such as HANARO and JRTR. The main drive mechanism of the BMCRDM is located in a Reactivity Control Mechanism (RCM) room under the reactor pool bottom, which makes the interference with equipment in the reactor pool reduced

Conceptual Design of Bottom-mounted Control Rod Drive Mechanism

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Haeng; Kim, Sanghaun; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Dongmin; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The arrangement of the BMCRDMs and irradiation holes in the core is therefore easier than that of the top-mounted CRDM. Hence, many foreign research reactors, such as JRR-3M, JMTR, OPAL, and CARR, have adopted the BMCRDM concept. The purpose of this paper is to introduce the basic design concept on the BMCRDM. The major differences of the CRDMs between HANARO and KJRR are compared, and the design features and individual system of the BMCRDM for the KJRR are described. The Control Rod Drive Mechanism (CRDM) is a device to regulate the reactor power by changing the position of a Control Absorber Rod (CAR) and to shut down the reactor by fully inserting the CAR into the core within a specified time. The Bottom-Mounted CRDM (BMCRDM) for the KiJang Research Reactor (KJRR) is a quite different design concept compared to the top-mounted CRDM such as HANARO and JRTR. The main drive mechanism of the BMCRDM is located in a Reactivity Control Mechanism (RCM) room under the reactor pool bottom, which makes the interference with equipment in the reactor pool reduced.

The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

Reversed field pinch reactor study 3

International Nuclear Information System (INIS)

Hollis, A.A.; Mitchell, J.T.D.

1977-12-01

This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

2003-01-01

The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

The World's Reactors no. 70 - Forsmark 3, BWR-75

International Nuclear Information System (INIS)

Anon.

1976-01-01

A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

International Nuclear Information System (INIS)

1979-11-01

This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

Successful BNCT for patients with cutaneous and mucosal melanomas. Report of 4 cases

International Nuclear Information System (INIS)

Morita, Norimasa; Hiratsuka, Junichi; Kuwabara, Chiaki; Aihara, Teruhito; Harada, Tamotsu; Imajo, Yoshinari; Ono, Koji; Fukuda, Hiroshi; Kumada, Hiroaki

2006-01-01

Since 2003 we have conducted BNCT clinical trials on melanomas at the Kyoto University Research Reactor (KUR) and Japan Research Reactor No.4 (JRR-4). We report 4 patients given BNCT for malignant melanomas: 2 with superficial spreading types on the heel, 1 with mucosal melanoma in the nasal cavity, and 1 with a melanoma on the vulva and in the vagina. The two cutaneous melanomas and the nasal cavity mucosal melanoma showed a complete response (CR) by 6 months after BNCT. The residual melanoma showed a partial response (PR) by 3 months after treatment and no regrowth since then. Although two patients experienced normal-tissue damage that exceeded the tolerance level, all the participants were cured within a few months of treatment. BNCT was shown to be a promising treatment for mucosal, as well as for cutaneous, melanomas. (author)

Development of telerobotic systems for reactor decommissioning, (3)

International Nuclear Information System (INIS)

Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

1991-01-01

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

Neutron diffractometer for bio-crystallography (BIX) with an imaging plate neutron detector

Energy Technology Data Exchange (ETDEWEB)

Niimura, Nobuo [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan)

1994-12-31

We have constructed a dedicated diffractometer for neutron crystallography in biology (BIX) on the JRR-3M reactor at JAERI (Japan Atomic Energy Research Institute). The diffraction intensity from a protein crystal is weaker than that from most inorganic materials. In order to overcome the intensity problem, an elastically bent silicon monochromator and a large area detector system were specially designed. A preliminary result of diffraction experiment using BIX has been reported. An imaging plate neutron detector has been developed and a feasibility experiment was carried out on BIX. Results are reported. An imaging plate neutron detector has been developed and a feasibility test was carried out using BIX.

A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

2007-11-01

The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

International Nuclear Information System (INIS)

El-Sahlamy, N.M.

2017-01-01

One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

The qualification of U3O8 as research reactor fuel

International Nuclear Information System (INIS)

Krull, W.

1983-01-01

This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

2004-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

International Nuclear Information System (INIS)

Khan, Jahirul Haque

2013-01-01

The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

International Nuclear Information System (INIS)

Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

1974-01-01

1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

A robot-automated work site for repair of the Chinon A3 reactor

International Nuclear Information System (INIS)

Raynal, A.

1987-01-01

In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time [fr

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

Decommissioning of the BR3 pressurized-water reactor

International Nuclear Information System (INIS)

Massaut, V.

1996-01-01

The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

Annual report of Nuclear Science Research Institute, JFY2006

International Nuclear Information System (INIS)

2008-03-01

Nuclear Science Research Institute (NSRI) is composed of Planning and Coordination Office and seven departments such as Department of Operational Safety Administration, Department of Radiation Protection, Department of Research Reactor and Tandem Accelerator, Department of Hot Laboratories and Facilities, Department of Criticality and Fuel Cycle Research Facilities, Department of Decommissioning and Waste Management, and Engineering Services Department. This annual report of JFY2006 summarizes the activities of NSRI, the R and D activities of the Research and Development Directorates and human resources development at site, and is expected to be referred to and utilized by R and D departments and project promotion sectors at NSRI site for the enhancement of their own research and management activities to attain their goals according to 'Middle-term Plan' successfully and effectively. In chapter 1, outline of JFY2006 activities of NSRI is described. In chapter 2, the following activities made by the departments in NSRI are summarized, i.e., (1) operation and maintenance of research reactors (JRR-3, JRR-4, NSRR), criticality assemblies (STACY, TRACY, FCA, TCA), hot laboratories (BECKY, Reactor Fuel Examination Facility, WASTEF, Research Laboratory 4, Plutonium Research Laboratory 1, Tokai Hot Laboratory, etc), and large-scale facilities (Tandem accelerator, LSTF, THYNC, TPTF, etc), and (2) safety management, radiation protection, management of radioactive wastes, decommissioning of nuclear facilities, engineering services, utilities and maintenance, etc, all of which are indispensable for the stable and safe operation and utilization of the research facilities. The technical developments for the advancement of the related technologies are also summarized. In chapter 3, the R and D and human resources development activities are described including the topics of the research works and projects performed by the Research and Development Directorates at site, such as

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Advances in Reactor physics, mathematics and computation. Volume 3

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.

1994-01-01

TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

On the major DYN3D developments for fast reactor design and transient analysis

International Nuclear Information System (INIS)

Merk, B.; Kliem, S.

2013-01-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

On the major DYN3D developments for fast reactor design and transient analysis

Energy Technology Data Exchange (ETDEWEB)

Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

2013-07-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija reaktora RA

Energy Technology Data Exchange (ETDEWEB)

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

1963-12-15

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report.

Operating reactors licensing actions summary. Vol. 3, No. 6

International Nuclear Information System (INIS)

1983-07-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Study of seismic responses of Candu-3 reactor building using isolator bearings

International Nuclear Information System (INIS)

Biswas, J.K.

1992-01-01

Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

Visualization of boron in molybdenum by α-rays track etching method and tritium autoradiography

International Nuclear Information System (INIS)

Saito, Hideo; Morita, Fumio

2003-01-01

Molybdenum alloys addicted with < 0.02 ppm B to 160 ppm B were analyzed by α-rays track etching (ATE) method irradiated by thermal neutron for 12 hours using atomic reactor of Rikkyo University and Japan atomic reactor of JRR-4. It was found that boron was segregated along grain boundaries and in the matrix. We analyzed boron distribution in the vicinity of the triple junctions at grain boundaries and in the matrix by the statistical frequency of α-rays tracks. Also we studied tritium autoradiography by cathodic charging method. Visualization of boron distribution was confirmed along the grain boundary which seemed to be effective trapping sites of hydrogen. (author)

Research on intelligent monitor for 3D power distribution of reactor core

International Nuclear Information System (INIS)

Xia, Hong; Li, Bin; Liu, Jianxin

2014-01-01

Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

Clinical practice in BNCT to the brain

International Nuclear Information System (INIS)

Nakagawa, Y.

2001-01-01

Our concept of Boron Neutron Capture Therapy (BNCT) is to selectively destroy tumour cells using the high LET particles yielded from the 10B(n,α)7Li reactions. The effort of clinical investigators has concentrated on how to escalate the radiation dose at the target point. BNCT in Japan combines thermal neutrons and BSH (Na 2 B 12 H 11 SH). The radiation dose is determined by the neutron fluence at the target point and the boron concentration in the tumour tissue. According to the recent analysis, the ratio of boron concentration (BSH) in tumour tissue and blood is nearly stable at around 1.2 to 1.69. Escalation of the radiation dose was carried out by means of improving the penetration of the thermal neutron beam. Since 1968, 175 patients with glioblastoma (n=83), anaplastic astrocytoma (n=44), low grade astrocytoma (n=16) or other types of tumour (n=32) were treated by BNCT at 5 reactors (HTR n=13, JRR-3 n=1, MulTR n=98, KUR n=30, JRR-2 n=33). The retrospective analysis revealed that the important factors related to the clinical results and QOL of the patients were minimum tumour volume radiation dose, more than 18Gy of physical dose and maximum vascular radiation dose (less than 15Gy) in the normal cortex. We have planned several trials to escalate the target radiation dose. One trial makes use of a cavity in the cortex following debulking surgery of the tumour tissue to improve neutron penetration. The other trial is introduction of epithermal neutron. KUR and JRR-4 were reconstructed and developed to be able to irradiate using epithermal neutrons. The new combination of surgical procedure and irradiation using epithermal neutrons should remarkably improve the target volume dose compared to the radiation dose treated by thermal neutrons. (author)

Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

International Nuclear Information System (INIS)

Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

2017-01-01

Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

Intraoperative boron neutron capture therapy for malignant gliomas. First clinical results of Tsukuba phase I/II trial using JAERI mixed thermal-epithermal beam

International Nuclear Information System (INIS)

Matsumura, A.; Yamamoto, T.; Shibata, Y.

2000-01-01

Since October 1999, a clinical trial of intraoperative boron neutron capture therapy (IOBNCT) is in progress at JRR-4 (Japan Research Reactor-4) in Japan Atomic Energy Research Institute (JAERI) using mixed thermal-epithermal beam (thermal neutron beam I: TNB-I). Compared to pure thermal beam (thermal neutron beam II: TNB-II), TNB-I has an improved neutron delivery into the deep region than TNB-II. The clinical protocol and the preliminary results will be discussed. (author)

Verification of dosimetry cross sections above 10 MeV based on measurement of activation reaction rates in fission neutron field

International Nuclear Information System (INIS)

Odano, Naoteru; Miura, Toshimasa; Yamaji, Akio.

1996-01-01

To validate the dosimetry cross sections in fast neutron energy range, activation reaction rates were measured for 5 types of dosimetry cross sections which have sensitivity in the energy rage above 10 MeV utilizing JRR-4 reactor of JAERI. The measured reaction rates were compared with the calculations reaction rates by a continuous energy monte carlo code MVP. The calculated reaction rates were based on two dosimetry files, JENDL Dosimetry File and IRDF-90.2. (author)

G2 and G3 reactors design; Description des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

1958-07-01

'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

Visualization of direct contact heat transfer between water and molten alloy by neutron radiography. 1

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi; Takenaka, Nobuyuki; Matsubayashi, Masahito.

1997-01-01

Design of an innovative Steam Generator (SG) for Liquid Metal Fast Reactors (LMFRs) using liquid-liquid direct contact heat transfer has been developing. In this concept, the SG shell is filled with a molten alloy, which is heated by primary sodium. Water is fed into the high-temperature, molten alloy, and evaporates by direct contact heating. In order to obtain the fundamental information needed to discuss the heat transfer mechanisms of direct contact between the water and molten alloy, this phenomenon was observed by neutron radiography. JRR-3M thermal neutron radiography at the Japan Atomic Energy Research Institute was used. This paper deals with the results of visualization of direct contact heat exchange in the molten alloy. (author)

The Texts of Instruments connected with the Agency's Supply of Uranium to Japan. Letters relating to the Agency's Safeguards and Health and Safety measures

International Nuclear Information System (INIS)

1962-01-01

On 2 March 1962 the Director General, pursuant to a decision taken by the Board of Governors on 28 February, addressed a letter to the Governor from Japan cancelling those provisions of the letter of 24 March 1959 that related to the safeguards against diversion applicable to the JRR-3 research reactor project. On 20 June the Director General, pursuant to a decision taken by the Board on 14 June, addressed a further letter to the Governor cancelling the remaining provisions of the letter of 24 March 1959, which related to health and safety measures for the project. The texts of the two letters in question are reproduced in this document for the information of all Members

3. International conference on catalysis in membrane reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

International Nuclear Information System (INIS)

Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

2003-01-01

Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

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

International Nuclear Information System (INIS)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun

1996-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Development of a 3-D flow analysis computer program for integral reactor

International Nuclear Information System (INIS)

Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

2003-01-01

A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

International Nuclear Information System (INIS)

Chi-Mei Lee; Kuo-Tsang Hung

2006-01-01

The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

International Nuclear Information System (INIS)

1992-04-01

Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

1992-01-01

The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

Annealing of the BR3 reactor pressure vessel

International Nuclear Information System (INIS)

Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

1985-01-01

The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

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

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

The challenge of observation on livings things by employing an ultra small-angle neutron scattering method

International Nuclear Information System (INIS)

Koizumi, Satoshi; Motokawa, Ryuhei; Iwase, Hiroki; Miyamoto, Nobuyoshi; Tanaka, Kazuhiro; Masui, Tomomi; Iida, You; Yue, Zhao; Chiba, Kaori; Kumada, Takayuki; Yamaguchi, Daisuke; Hashimoto, Takeji

2007-01-01

To address the question as to how small-angle scattering is effectively applied to the cell, i.e., a hierarchically ordered system comprising multi-components of macro and small molecules, the size of which ranges from 100 μm to several μm, we reconstructed SANS-J (pinhole small-angle neutron scattering spectrometer at research reactor JRR3, Tokai) to focusing and polarized neutron small-angle spectrometer (SANS-J-II), by employing focusing neutron lenses and high resolution photomultiplier. Consequently, an accessible minimum wave number q min was improved from 3x10 -3 A -1 to medium ultra-small angle scattering of 3x10 -4 A -1 . The focusing USANS method, thus developed, is crucial to fill the gap in wave number q between those covered by a double crystal method and by a conventional pin-hole method. (author)

Master-3.0: multi-purpose analyzer for static and transient effects of reactors

International Nuclear Information System (INIS)

Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

2002-03-01

MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

3 D flow computations under a reactor vessel closure head

International Nuclear Information System (INIS)

Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

1995-12-01

The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

Production of radioisotopic gamma radiation sources in JAERI

International Nuclear Information System (INIS)

Katoh, Hisashi; Kogure, Hiroto; Suzuki, Kyohei

1980-04-01

The present state of production of gamma radiation sources in Japan Atomic Energy Research Institute (JAERI) is described. Sources of 192 Ir, 60 Co and 170 Tm for industrial and 198 Au and 192 Ir for medical applications are produced and delivered routinely by JAERI. Prefabricated assembly targets are irradiated in JRR-2, JRR-3, JRR-4 or JMTR. The irradiated targets are disassembled in a heavy density concrete cave or a lead-shielded cell, depending on the level of radioactivity. The yield of radioactivity in each target is measured with the aid of an ionization chamber. Where necessary, irradiated targets are encapsulated hermetically in capsules of aluminium, stainless steel or other material. The yield of radioactivity is estimated in relation with the burn-up of target nuclide and product nuclide. (author)

Irradiation effects on C/C composite materials for high temperature nuclear applications

International Nuclear Information System (INIS)

Eto, M.; Ugachi, H.; Baba, S.I.; Ishiyama, S.; Ishihara, M.; Hayashi, K.

2000-01-01

Excellent characteristics such as high strength and high thermal shock resistance of C/C composite materials have led us to try to apply them to the high temperature components in nuclear facilities. Such components include the armour tile of the first wall and divertor of fusion reactor and the elements of control rod for the use in HTGR. One of the most important aspects to be clarified about C/C composites for nuclear applications is the effect of neutron irradiation on their properties. At the Japan Atomic Energy Research Institute (JAERI), research on the irradiation effects on various properties of C/C composite materials has been carried out using fission reactors (JRR-3, JMTR), accelerators (TANDEM, TIARA) and the Fusion Neutronics Source (FNS). Additionally, strength tests of some neutron-irradiated elements for the control rod were carried out to investigate the feasibility of C/C composites. The paper summarises the R and D activities on the irradiation effects on C/C composites. (authors)

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

3D CFD for chemical transport profiles in a rotating disk CVD reactor

Science.gov (United States)

Han, Jong-Hyun; Yoon, Do-Young

2010-06-01

The RDCVD (Rotating Disk Chemical Vapor Deposition) technique is an appropriate method for uniform deposition of grains, such as compound semiconductior materials. The substrate temperature and rotation speed are the major factors, which determine the thickness uniformity of the deposited films. This paper investigates 3D CFD (3 Dimensional Computational Fluid Dynamics) simulation results of flow and heat transfer in a reactor of RDCVD using Fluent. In order to establish the reducibility of buoyancy effect on deposition quality, the chemical transport profile upon the disk heated is examined successfully in 3D domain for different rotating speeds. The resulting vortex flows due the simultaneous buoyance and centrifuge are discussed qualitatively in the 3D virtual system of a RDCVD reactor. 3D CFD is even more effective to describe the internal vortex flows due to the competitive inlet, buoyancy and centrifuge flows, which cannot be realized in the general 2D (2 Dimensional) CFD.[Figure not available: see fulltext.

Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

2012-07-15

Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

KAUST Repository

Karsenty, Florent

2012-08-16

Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

International Nuclear Information System (INIS)

Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

2015-01-01

Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

Study on effects of development of reactor constant in fast reactor analysis

International Nuclear Information System (INIS)

Chiba, Gou

2002-12-01

Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

Energy Technology Data Exchange (ETDEWEB)

Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

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

Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

International Nuclear Information System (INIS)

Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

2017-01-01

Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Tritium production, management and its impact on safety for a D-3He fusion reactor

International Nuclear Information System (INIS)

Sze, D.K.; Herring, S.; Sawan, M.

1991-11-01

About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

Small-angle neutron scattering instrument of Institute for Solid State Physics, the Univeristy of Tokyo (SANS-U) and its application to biology

Energy Technology Data Exchange (ETDEWEB)

Ito, Yuji; Imai, Masayuki; Takahashi, Shiro [Univ. of Tokyo, Tokai Naka Ibaraki (Japan)

1994-12-31

A small-angle neutron spectrometer (SANS-U) suitable for the study of mesoscopic structure in the field of polymer chemistry and biology, has been constructed at the guide hall of JRR-3M reactor at the Japan Atomic Energy Research Institute. The instrument is 32m long and utilizes a mechanical velocity selector and pinhole collimation to provide a continuous beam with variable wavelength in the range from 5 to 10{Angstrom}. The neutron detector is a 65 x 65cm{sup 2} 2D position sensitive proportional counter. The practical Q range of SANS-U is 0.0008 to 0.45{Angstrom}{sup -1}. The design, characteristics and performance of SANS-U are described with some biological studies using SANS-U.

Characteristics of UV-MicroO3 Reactor and Its Application to Microcystins Degradation during Surface Water Treatment

Directory of Open Access Journals (Sweden)

Guangcan Zhu

2015-01-01

Full Text Available The UV-ozone (UV-O3 process is not widely applied in wastewater and potable water treatment partly for the relatively high cost since complicated UV radiation and ozone generating systems are utilized. The UV-microozone (UV-microO3, a new advanced process that can solve the abovementioned problems, was introduced in this study. The effects of air flux, air pressure, and air humidity on generation and concentration of O3 in UV-microO3 reactor were investigated. The utilization of this UV-microO3 reactor in microcystins (MCs degradation was also carried out. Experimental results indicated that the optimum air flux in the reactor equipped with 37 mm diameter quartz tube was determined to be 18∼25 L/h for efficient O3 generation. The air pressure and humidity in UV-microO3 reactor should be low enough in order to get optimum O3 output. Moreover, microcystin-RR, YR, and LR (MC-RR, MC-YR, and MC-LR could be degraded effectively by UV-microO3 process. The degradation of different MCs was characterized by first-order reaction kinetics. The pseudofirst-order kinetic constants for MC-RR, MC-YR, and MC-LR degradation were 0.0093, 0.0215, and 0.0286 min−1, respectively. Glucose had no influence on MC degradation through UV-microO3. The UV-microO3 process is hence recommended as a suitable advanced treatment method for dissolved MCs degradation.

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

International Nuclear Information System (INIS)

Reyes F, M. C.; Del Valle G, E.; Filio L, C.

2013-10-01

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

Annual report on operation, utilization and technical development of hot laboratories. From April 1, 1996 to March 31, 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

This report describes activities, in fiscal year 1996, of the Reactor Fuel Examination Facility (RFEF), the Research Hot Laboratory (RHL) and the Waste Safety Testing Facility (WASTEF) which belong to the Department of Hot laboratories. In the RFEF, Post-Irradiation Examinations (PIEs) of PWR fuel assemblies irradiated in the Takahama Unit 3, a BWR fuel assembly irradiated in the Fukusima Daini Unit have been performed. Also, PIEs of assembly materials irradiated in the Fugen Reactor have been carried out. To support R and D works in JAERI, refabrication of segmented fuel rods have been done using irradiated LWR fuel rods for pulse irradiation in the NSRR and re-irradiation tests in the JMTR. PIEs have been performed on high burnup fuel rods and ROX fuel rods. For the RHL, PIEs have been performed on segment fuels irradiated in the NSRR, fuels and materials for HTTR, standard fuels for JRR-3M and materials for nuclear fusion reactor. In addition, a monitoring test of fuel elements in accordance with the surveillance program of the Magnox reactor of the Japan Atomic Power Corporation has been continued. In the WASTEF, leaching tests on TRU in simulated glass forms and a low flow rate tests on glass waste forms have been carried out. The examinations of alpha damage acceleration for the Synroc waste forms have also been performed. (author)

Quantitative analyses of impurity silicon-carbide (SiC) and high-purity-titanium by neutron activation analyses based on k0-standardization method. Development of irradiation silicon technology in productivity using research reactor (Joint research)

International Nuclear Information System (INIS)

Motohashi, Jun; Takahashi, Hiroyuki; Magome, Hirokatsu; Sasajima, Fumio; Tokunaga, Okihiro; Kawasaki, Kozo; Onizawa, Koji; Isshiki, Masahiko

2009-07-01

JRR-3 and JRR-4 have been providing neutron-transmutation-doped silicon (NTD-Si) by using the silicon NTD process, which is a method to produce a high quality semiconductor. The domestic supply of NTD-Si is insufficient for the demand, and the market of NTD-Si is significantly growing at present. It is very important to increase achieve the production. To fulfill the requirement, we have been investigating a neutron filter, which is made of high-purity-titanium, for uniform doping. Silicon-carbide (SiC) semiconductor doped with NTD technology is considered suitable for high power devices with superior performances to conventional Si-based devices. We are very interested in the SiC as well. This report presents the results obtained after the impurity contents in the high-purity-titanium and SiC were analyzed by neutron activation analyses (NAA) using k 0 -standardization method. There were 6 and 9 impurity elements detected from the high-purity-titanium and SiC, respectively. Among those Sc from the high-purity-titanium and Fe from SiC were comparatively long half life nuclides. From the viewpoint of exposure in handling them, we need to examine the impurity control of materials. (author)

Annual report of Nuclear Science Research Institute, JFY2005

International Nuclear Information System (INIS)

2007-04-01

Japan Atomic Energy Agency (JAEA) was inaugurated on October 1st, 2005. Works for the operation and maintenance of various research facilities as well as safety management, radiation protection, and radioactive wastes management, which have been undertaken by departments in Tokai Research Establishment of Japan Atomic Energy Research Institute (JAERI), were inherited by newly established departments of Nuclear Science Research Institute (NSRI). The NSRI is composed of Planning and Coordination Office and seven departments such as Department of Operational Safety Administration, Department of Radiation Protection, Department of Research Reactor and Tandem Accelerator, Department of Hot Laboratories and Facilities, Department of Criticality and Fuel Cycle Research Facilities, Department of Decommissioning and Waste Management, and Engineering Services Department. This annual report of JFY 2005 summarizes the activities of NSRI and is expected to be referred to and utilized by R and D departments and project promotion sectors at NSRI site for the enhancement of their own research and management activities to attain their goals according to Middle-term Plan' successfully and effectively. In chapter 1, outline of organization and administrative activities of NSRI is described. In chapter 2, the following activities made by the departments in NSRI are summarized, i.e., (1) operation and maintenance of research reactors (JRR-3, JRR-4, NSRR), criticality assemblies (STACY, TRACY, FCA, TCA), hot laboratories, (BECKY, Reactor Fuel Examination Facility, WASTEF, Research Laboratory 4, Plutonium Research Laboratory 1, Tokai Hot Laboratory, etc), and large-scale facilities (Tandem accelerator, LSTF, THYNC, TPTF, etc), and (2) safety management, radiation protection, management of radioactive wastes, decommissioning of nuclear facilities, engineering services, utilities and maintenance, etc, all of which are indispensable for the stable and safe operation and utilization of the

Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

2008-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

International Nuclear Information System (INIS)

Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

2013-01-01

Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

International Nuclear Information System (INIS)

Greene, S.R.; Fisher, S.E.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

Fuel safety research 1999

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-07-01

In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

A 3D transport-based core analysis code for research reactors with unstructured geometry

International Nuclear Information System (INIS)

Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

2013-01-01

Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

Plutonium-burn high temperature gas-cooled reactor for 3E+3S

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst

Directory of Open Access Journals (Sweden)

New Pei Yee

2008-04-01

Full Text Available A one-dimensional mathematical model was developed to simulate the performance of catalytic fixed bedreactor for carbon dioxide reforming of methane over Rh/Al2O3 catalyst at atmospheric pressure. The reactionsinvolved in the system are carbon dioxide reforming of methane (CORM and reverse water gas shiftreaction (RWGS. The profiles of CH4 and CO2 conversions, CO and H2 yields, molar flow rate and molefraction of all species as well as reactor temperature along the axial bed of catalyst were simulated. In addition,the effects of different reactor temperature on the reactor performance were also studied. The modelscan also be applied to analyze the performances of lab-scale micro reactor as well as pilot-plant scale reactorwith certain modifications and model verification with experimental data. © 2008 BCREC UNDIP. All rights reserved.[Received: 20 August 2008; Accepted: 25 September 2008][How to Cite: N.A.S. Amin, I. Istadi, N.P. Yee. (2008. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst. Bulletin of Chemical Reaction Engineering and Catalysis, 3 (1-3: 21-29. doi:10.9767/bcrec.3.1-3.19.21-29

A new irradiation method with a neutron filter for silicon neutron transmutation doping at the Japan research reactor no. 3 (JRR-3).

Science.gov (United States)

Komeda, Masao; Kawasaki, Kozo; Obara, Toru

2013-04-01

We studied a new silicon irradiation holder with a neutron filter designed to make the vertical neutron flux profile uniform. Since an irradiation holder has to be made of a low activation material, we applied aluminum blended with B4C as the holder material. Irradiation methods to achieve uniform flux with a filter are discussed using Monte-Carlo calculation code MVP. Validation of the use of the MVP code for the holder's analyses is also discussed via characteristic experiments. Copyright © 2013 Elsevier Ltd. All rights reserved.

Annual report of the Japan Atomic Energy Research Institute, for fiscal 1989

International Nuclear Information System (INIS)

1990-01-01

Japan Atomic Energy Research Institute has promoted the research on nuclear safety, the research and development of high temperature engineering and nuclear fusion which are the leading projects bringing about the breakthrough in atomic energy technology, the research on radiation utilization and the research and development of nuclear-powered ships, following the 'Plan of development and long term utilization of atomic energy' decided in 1987, as the central, general research institute in atomic energy field in Japan. Also the advanced basic research for opening atomic energy frontier and various international cooperation as well as the cooperation in Japan have been promoted. The engineering safety of nuclear facilities and environmental safety, the construction of the Nuclear Fuel Cycle Safety Engineering Research Facility, the design of the High Temperature Engineering Test Reactor and the various tests related to it, the reconstruction of JT-60 for increasing the current, the design of a nuclear fusion reactor, the high utilization of radiation using ion beam, the construction of Sekinehama Port for the nuclear-powered ship 'Mutsu', the power increasing test of the reactor of the Mutsu, the reconstruction of JRR-3 and others are reported. (K.I.)

Collective occupational dose for nuclear reactors of the 2., 3. and 4. generation

International Nuclear Information System (INIS)

Guidez, J.; Saturnin, A.

2016-01-01

In France during reactor operation the individual occupational doses are collected and recorded according to the law. When you sum up all the individual doses you get the yearly collective dose expressed in Man.Sv/year. This piece of information can be used to make comparisons between various types of reactors and between reactors of the same type. The results show a steady decrease of the collective dose for all types of reactors over the time except for CANDU reactors for which a slight increase of the dose has appeared since the years 1996-1998. The decrease is due to the continuous improvement of reactor operating and to changes in the reactor design. There is also a constant gap over time between the collective dose for a BWR reactor (1.12 Man.Sv/y) and a PWR reactor 0.60 Man.Sv/y), this gap is certainly due to N 16 nuclide that is created in the primary circuit and transported to turbines in the case of a BWR reactor. For sodium-cooled fast reactors (RNR-Na) the collective dose is below 0.40 Man.Sv/y except for the BN-600 reactor. (A.C.)

RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

International Nuclear Information System (INIS)

Zecevic, V.

1963-12-01

This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

Report on progress of researches by common utilization of JAERI nuclear facilities, for fiscal, 1991

International Nuclear Information System (INIS)

1992-08-01

The results of the joint researches by utilizing the facilities of JAERI in 1991 fiscal year were summarized, and this report was able to be completed. Many researchers in whole Japan took part in many themes, and the very significant results were obtained. Now this joint research has reached the great turnabout period. The reconstructed JRR-3M was offered for joint utilization since April, 1991, and the utilization for neutron diffraction and scattering increased largely. As for the ion irradiation facility in Takasaki Research Establishment, the partial operation will be started in the next year, and the joint utilization is expected to begin. Accompanying the diversification of the utilization of facilities, in order to properly meet the needs of users, the thorough revision of the system seems necessary. The number of research themes in 1991 was 222 cases. JRR-3M accomplished the joint utilization operation of 8 cycles as expected, but JRR-2 caused a trouble during 5th cycle, and the operation thereafter was canceled. In this book, 159 reports are collected. (K.I.)

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

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)

Fuel safety research 2001

Energy Technology Data Exchange (ETDEWEB)

Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-11-01

The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

Reactor safety study applied to the Forsmark 3 Power Plant

International Nuclear Information System (INIS)

Ericsson, G.; Tiren, L.I.

1978-01-01

A reactor safety study of the Forsmark 3 BWR power plant has been carried out for the purpose of calculating core melt probabilities using WASH-1400 methods. A sensitivity analysis shows that the calculated core melt probability is changed by approximately a factor of 10 depending on assumptions made with respect to the probability of human error. The importance of the availability of off-site power and the influence of common cause failure is also discussed. (author)

Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

International Nuclear Information System (INIS)

Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

2009-01-01

A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

Development of a 3-dimensional calculation model of the Danish research reactor DR3 to analyse a proposal to a new core design called ring-core

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E

1985-07-01

A 3-dimensional calculation model of the Danish research reactor DR3 has been developed. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. The model has been tested against measurements with satisfying results. Furthermore the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dummy elements to increase the thermal flux in the center of the reactor. (author)

The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

1997-01-01

The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

International Nuclear Information System (INIS)

Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

2005-01-01

Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

Revision of fast reactor group constant set JFS-3-J2

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

Verification of the CASMO-3/SIMULATE-3 pin power accuracy by comparison with operating boiling water reactor measurements

International Nuclear Information System (INIS)

Uegata, T.; Saji, E.; Tanaka, H.

1993-01-01

Intranodal pin power distributions calculated by the CASMO-3/SIMULATE-3 code have been compared with pin gamma scan measurements. These data were obtained from the depleted core of an operating boiling water reactor (BWR), which is more complicated than a pressurized water reactor to calculate because of the existence of coolant void distributions and cruciform control blades. Furthermore, measured bundles include mixed-oxide (MOX) bundles in which steep thermal flux gradients occur. Both UO 2 and MOX bundles have been calculated in the same manner based on the standard CASMO-3/SIMULATE-3 methods. The total pin power root-mean-square (rms) error is 2.7%, which includes measurement error, from an 896-point comparison. There is no obvious dependency on axial elevations (void fractions) and no significant difference between fuel types (UO 2 or MOX), although the errors in a peripheral bundle, which is less important from the standpoint of core design, are somewhat larger than those in the internal bundles. If the peripheral bundle is excluded, the total rms error is reduced to 2.2%. From these results, it is concluded that excellent agreement has been obtained between the calculations and measurements and that the calculational capability of CASMO-3/SIMULATE-3 for the intranodal pin power distribution is quite satisfactory and useful for BWR core design

Generation IV reactors: economics

International Nuclear Information System (INIS)

Dupraz, B.; Bertel, E.

2003-01-01

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

Status of French reactors

International Nuclear Information System (INIS)

Ballagny, A.

1997-01-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm 3 . The OSIRIS reactor has already been converted to LEU. It will use U 3 Si 2 as soon as its present stock of UO 2 fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU

Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

International Nuclear Information System (INIS)

Tuunanen, J.

1993-01-01

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

Energy Technology Data Exchange (ETDEWEB)

C. B. Davis

2006-07-01

The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Radiation protection at the RA Reactor in 1993, RA research reactor, Part

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

1993-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry and radiation protection at the RA reactor; (2) decontamination, collecting and treatment of fluid effluents and solid wastes; (3) Radioactivity control in the vicinity of the reactor and (4)meteorology measurements; (3). Each of the category is described as a separate annex of this report [sr

Analysis of a total flow blockage of a Fuel Assembly in a typical MTR Research Reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Adorni, M.; Salah, A.B.; Di Maro, B.; Pierro, F.; D'Auria, F.; Hamidouche, T.

2004-01-01

The lack of full understanding of complex mechanisms connected with the interaction between thermal-hydraulics and neutronics still challenge the design and the operation of nuclear reactors by the adoption of conservative safety limits. The recent availability of powerful computer and computational techniques together with the continuing increase in operational experience imposes the revisiting of those areas and eventually the identification of design/safety requirements that can be relaxed [1]. Currently, the enlarged commercial exploitation of nuclear Research Reactors (RR) has increased the consideration to their corresponding safety issues. Almost all of the safety analyses have so far been performed using conservative computational tools [2]. Nowadays, the application of Best-Estimate (BE) methods constitutes a real necessity in order to increase their commercial productivity. In this framework, an attempt is made to apply the BE technique to perform a safety evaluation under research reactors operational conditions. In fact, this technique has been largely verified and validated for power reactors using coupled system thermal-hydraulic and three-dimensional neutron kinetics [1]. For this purpose, as typical representative of research reactors, the IAEA 10 MW MTR Research Reactors problem [3] is considered. The system thermal-hydraulic RELAP5 [4] code was developed to simulate transient scenarios in Power reactors such PWR, BWR, VVER, etc. However, only limited work was performed to access the applicability of the code to Research Reactors operating conditions (low pressure, mass flow rates, power, etc) [5]. Previous works performed in this field are reported in [5], [6] and [7]. In this framework, total and partial blockage of a single Fuel Assembly cooling channel are investigated. As a first attempt the calculations are performed by applying the BE thermal-hydraulic system code RELAP5 alone using its point kinetic model to derive the instantaneous core

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program

International Nuclear Information System (INIS)

Hernandez, N.; Alonso, G.; Valle, E. del

2004-01-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

Verification of the computational dosimetry system in JAERI (JCDS) for boron neutron capture therapy

International Nuclear Information System (INIS)

Kumada, H; Yamamoto, K; Matsumura, A; Yamamoto, T; Nakagawa, Y; Nakai, K; Kageji, T

2004-01-01

Clinical trials for boron neutron capture therapy (BNCT) by using the medical irradiation facility installed in Japan Research Reactor No. 4 (JRR-4) at Japan Atomic Energy Research Institute (JAERI) have been performed since 1999. To carry out the BNCT procedure based on proper treatment planning and its precise implementation, the JAERI computational dosimetry system (JCDS) which is applicable to dose planning has been developed in JAERI. The aim of this study was to verify the performance of JCDS. The experimental data with a cylindrical water phantom were compared with the calculation results using JCDS. Data of measurements obtained from IOBNCT cases at JRR-4 were also compared with retrospective evaluation data with JCDS. In comparison with phantom experiments, the calculations and the measurements for thermal neutron flux and gamma-ray dose were in a good agreement, except at the surface of the phantom. Against the measurements of clinical cases, the discrepancy of JCDS's calculations was approximately 10%. These basic and clinical verifications demonstrated that JCDS has enough performance for the BNCT dosimetry. Further investigations are recommended for precise dose distribution and faster calculation environment

Survey of group data libraries for use of the DYN3D program for WWER type reactors

International Nuclear Information System (INIS)

Mittag, S.

1994-06-01

So-called few-group neutron data have to be used as input data in core models (such as DYN3D) calculating the reactor behaviour. A survey is given of qualified data libraries for the reactor cores of Russian VVER. The information about primary data used in group data generation and the accuracy reached by the cell codes is compiled in tables. To assess the quality of the data, comparisons have been made between measured and calculated reactor parameters. The information available does not show significant differences concerning the quality of the data libraries. (orig.) [de

Status of French reactors

Energy Technology Data Exchange (ETDEWEB)

Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

Validation and application of 3D-methods for the design and safety analysis of high temperature reactors

International Nuclear Information System (INIS)

Bader, J.; Lapins, J.; Buck, M; Bernnat, W.; Laurien, E.

2011-01-01

Some of the concepts for future nuclear reactors are high-temperature gas-cooled reactors. Previous simulation codes for their cores were often based on one- or two-dimensional models, but today's increasing computer capabilities make an advance to 3D-codes possible now. Our thermal-hydraulic code ATTICA3D (Advanced Thermal-hydraulic Tool for In-vessel and Core Analysis in 3 Dimensions) is based on the porous media approach, including 3-D models of heat conduction and gas flow, using a coarse-grid integration method for the time-dependent conservation equations of mass, momentum and energy. Results of numerical calculations for various validation cases are presented: First, the test facility SANA is chosen, which has been used to study heat transfer phenomena inside a coolant-gas filled pebble-bed core, which was heated by embedded electrical heating elements. Calculations were carried out for different tests taken from the experimental database. Measured and calculated temperatures at different positions are compared and found in good agreement. Second, our code was used to simulate a depressurized loss of forced cooling experiment with simulated decay heat in the AVR Experimental Reactor. Due to its design with the shut-down rods located inside columnar noses, which extend into the pebble bed of the core, geometry and power distribution are genuinely three-dimensional. The power distribution was calculated by the 3D-Neutronic Diffusion Code CITATION in conjunction with the spectral code MICROX-2. The neutronics and thermal-hydraulics calculations were carried out for a 3D, 45°-degree section of the reactor. It is demonstrated, that the experimental results could be qualitatively reproduced. (author)

VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 3. Programmer's manual. Final report

International Nuclear Information System (INIS)

Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

1983-05-01

VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear-reactor-core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This is Volume 3, the Programmer's Manual. It explains the codes' structures and the computer interfaces

Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-01

This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor.

Atomization of U3Si2 for research reactor fuel

International Nuclear Information System (INIS)

Kim, C.K.; Kim, K.H.; Lee, C.T.; Kuk, I.H.

1995-01-01

Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U 3 Si and the mixture of U 3 Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U 3 Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

Basic research of neutron radiography using cold neutron beam

International Nuclear Information System (INIS)

Oda, Masahiro; Tamaki, Masayoshi; Tasaka, Kanji

1995-01-01

As the result of demanding high quality images, now the nuclear reactors which can supply stably intense neutron beam have become the most general neutron source for radiography. For the purpose, mostly thermal neutrons have been used, but it is indispensable to use other neutrons than thermal neutrons for advancing neutron radiography technology and expanding the application fields. The radiography using cold neutrons is most behind in the development because the suitable neutron source was not available in Japan. The neutron sources for exclusively obtaining intense cold neutron beam were installed in the Kyoto University reactor in 1986 and in the JRR-3M of Japan Atomic Energy Research Institute in 1991. Basically as neutron energy lowers, the cross section of substances increases. In certain crystalline substances, the Bragg cutoff arises. The removal of scattered neutrons, the measurement of parallelism of beam and the relation of the thickness of objects with the transmissivity of cold neutrons are described. The imaging by TV method and the cold neutron CT in the CNRF and the simplified neutron CT by film method are reported. (K.I.)

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

Science.gov (United States)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

Advances in reactor physics education: Visualization of reactor parameters

International Nuclear Information System (INIS)

Snoj, L.; Kromar, M.; Zerovnik, G.

2012-01-01

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

DRAGON 3.05D, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: The computer code DRAGON contains a collection of models that can simulate the neutron behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. 2 - Methods: The code DRAGON contains a multigroup flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries. The execution of DRAGON is

Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

1998-03-01

The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

Detailed modeling of KALININ-3 NPP VVER-1000 reactor pressure vessel by the coupled system code ATHLET/BIPR-VVER

International Nuclear Information System (INIS)

Nikonov, S.P.; Velkov, K.; Pautz, A.

2011-01-01

The paper gives an overview of the recent developments of a new reactor pressure vessel (RPV) model of VVER-1000 for the coupled system code ATHLET/BIPR-VVER. Based on the previous experience a methodology is worked out for modeling the RPV in a pseudo-3D way with the help of a multiple parallel thermal-hydraulic channel scheme that follows the hexagonal fuel assembly structure from the bottom to the top of the reactor. The results of the first application of the new modeling are discussed on the base of the OECD/NEA coupled code benchmark for Kalinin-3 NPP transient. Coolant mass flow distributions in reactor volume of VVER 1000 reactor are presented and discussed. It is shown that along the core height a mass flow re-distribution of the coolant takes place starting approximately at an axial layer located 1 meter below the core outlet. (author)

Comparative studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 data libraries on the Monte Carlo continuous energy modeling of the gas turbine-modular helium reactor operating with thorium fuels

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile 232 Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of 239 Pu, 233 U and 235 U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B. (author)

3 MW TRIGA Research Reactor facility of BAEC and its Utilization

International Nuclear Information System (INIS)

Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z.

1999-01-01

The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

Verification of RBMK-1500 reactor main circulation circuit model with Cathare V1.3L

International Nuclear Information System (INIS)

Jasiulevicius, A.

2001-01-01

Among other computer codes, French code CATHARE is also applied for RBMK reactor calculations. In this paper results of such application for Ignalina NPP reactor (RBMK-1500 type) main circulation circuit are presented. Three transients calculations were performed: all main circulation pumps (MCP) trip, trip of one main circulation pump and trip of one main circulation pump without a closure of check valve on the pump line. Calculation results were compared to data from the Ignalina NPP, where all these transients were recorded in the years 1986, 1996 and 1998. The presented studies prove the capability of the CATHARE code to treat thermal-hydraulic transients with a reactor scram in the RBMK, in case of single or multiple pump trips. However, the presented model needs further improvements in order to simulate loss of coolant accidents. For this reason, emergency core cooling system should be included in the model. Additional model improvement is also needed in order to gain more independent pressure behavior in both loops. Also, flow rates through the reactor channels should be modeled by dividing channels into several groups, referring to channel power (in RBMK power produced in a channel, located in different parts of the core is not the same). The point-neutron kinetic model of the CATHARE code is not suitable to predict transients when the reactor is operating at a nominal power level. Such transients would require the use of 3D-neutron kinetics model to describe properly the strong space-time effect on the power distribution in the reactor core