Site preparation and excavation works for the foundation of station main building among construction works for No. 1 unit in Kashiwazaki-Kariwa Nuclear Pwer Station

International Nuclear Information System (INIS)

Ueyama, Koreyasu

1982-01-01

Tokyo Electric Power Co., Inc., is planning the nuclear power station of final capacity 8,000 MW (7 units) in the region spread over Kashiwazaki City and Kariwa Village in Niigata Prefecture. For No. 1 unit (1100 MWe BWR), the reactor installation license was obtained in September, 1977, the site preparation and road construction started in April, 1978, and harbour construction works started in August, 1979. The construction works are now at the peak, and the overall progressing rate as of the end of June, 1982, is about 51 %. The site is a hilly region of dune along the coast of the Sea of Japan, and No. 1 unit is located in the southern part of the site. This paper reports on the outline of the project, site preparation and excavation works for the foundation of the station main building. For the site preparation and the excavation works for the foundation the main building, the shape of slope cutting, the design of landslide-preventing wall for the vertical excavation for the reactor complex building, and the construction plan and the result are reported. For underground water impermeable wall works, its outline, groundwater condition, groundwater simulation analysis, the investigation of wall installation, the wall structure and construction are described in detail. Also the outline of the control of slope face measurement, the control standards and the measured results are reported. (Wakatsuki, Y.)

Site preparation and excavation works for the foundation of station main building among construction works for No. 1 unit in Kashiwazaki-Kariwa Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

Ueyama, Koreyasu [Tokyo Electric Power Co., Inc. (Japan)

1982-09-01

Tokyo Electric Power Co., Inc., is planning the nuclear power station of final capacity 8,000 MW (7 units) in the region spread over Kashiwazaki City and Kariwa Village in Niigata Prefecture. For No. 1 unit (1100 MWe BWR), the reactor installation license was obtained in September, 1977, the site preparation and road construction started in April, 1978, and harbour construction works started in August, 1979. The construction works are now at the peak, and the overall progressing rate as of the end of June, 1982, is about 51 %. The site is a hilly region of dune along the coast of the Sea of Japan, and No. 1 unit is located in the southern part of the site. This paper reports on the outline of the project, site preparation and excavation works for the foundation of the station main building. For the site preparation and the excavation works for the foundation the main building, the shape of slope cutting, the design of landslide-preventing wall for the vertical excavation for the reactor complex building, and the construction plan and the result are reported. For underground water impermeable wall works, its outline, groundwater condition, groundwater simulation analysis, the investigation of wall installation, the wall structure and construction are described in detail. Also the outline of the control of slope face measurement, the control standards and the measured results are reported.

Preliminary findings and lessons learned from the 16 July 2007 earthquake at Kashiwazaki-Kariwa NPP- 'The Niigataken Chuetsu-Oki earthquake, Kashiwazaki-Kariwa NPP and Tokyo, Japan, 6-10 August 2007. Mission report. V. 1[Report to the Government of Japan

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-08-17

Upon request from the Government of Japan an IAEA expert mission was conducted at the Kashiwazaki-Kariwa NPP following a strong earthquake that affected the plant on 16 July 2007. The objective, as agreed with the Japanese counterpart, was to conduct a fact finding mission and to identify the preliminary lessons learned that might have implications for the international nuclear safety regime. Although the Niigataken Chuetsu-Oki earthquake on 16 July 2007 significantly exceeded the level of the seismic input taken into account in the design of the plant, the installation behaved in a safe manner, during and after the earthquake. In particular, the automatic shutdown of the reactors of Units 3, 4 and 7, which were operating at full power, and of the reactor of Unit 2, which was in the start up state, were performed successfully. Based on the reports from experts from the Tokyo Electric Power Company (TEPCO) and the limited but representative plant walkdowns and visual observations performed by the IAEA team, safety related structures, systems and components of the plant seem to be in a much better general condition than might be expected for such a strong earthquake, and there is no visible significant damage. This is probably due to the conservatisms introduced at different stages of the design process. The combined effects of these conservatisms were apparently sufficient to compensate for uncertainties in the data and methods available at the time of the design of the plant, which led to the underestimation of the original seismic input. However, important components like the reactor vessels, the core internals and the fuel elements have not yet been examined and in-depth inspections are still to be performed. On the other hand, non-safety related structures, systems and components were affected by significant damage such as soil and anchorage failures and oil leakages. A re-evaluation of the seismic safety the Kashiwazika-Kariwa NPP needs to be done with account

Economic and financial benefits as a compensation for living near a nuclear power station. A case study of Kashiwazaki-Kariwa Nuclear Power Station

International Nuclear Information System (INIS)

Kato, Takaaki; Hatta, Masahisa; Matsumoto, Shiro; Nishikawa, Masashi

2007-01-01

Although dwellers living near a nuclear power station are entitled to economic/financial benefits such as increased job opportunities and local tax revenues pertaining to the power station, it is not clear whether such benefits are appreciated by the dwellers. Two findings of this study based upon a social survey of local dwellers living near the Kashiwazaki-Kariwa Nuclear Power Station are summarized as follows. First, an increase in the per capita sizes of the local tax revenue and national subsidies resulted in a larger share of respondents who thought that those revenues are beneficial. Therefore, local dwellers are aware of the sizes of economic/financial benefits. Second, given the same risk level of nuclear disaster, a larger per capita financial benefit resulted in a larger share of respondents who felt compensated for the nuclear risk. However, this increase in the number of compensated respondents is low relative to the increase in the amount of financial benefits. (author)

Advanced boiling water reactor (ABWR). Design, construction, operation and maintenance experience

International Nuclear Information System (INIS)

Idesawa, M.

1998-01-01

The ABWR has experienced all phases of design, construction, operation and maintenance at Kashiwazaki-Kariwa Nuclear Power Station Units No.6 and 7 and confirmed that originally intended development targets have been achieved with highly satisfactory results. This is the fruit of a project that collected wisdom from various sources under a international cooperative organization, with Tokyo Electric Power Company taking the leading role from the onset. These two units have not only demonstrated that ABWRs have superior performance as the first standard units of advanced light water reactor but also aroused a hope for the big potential advantages that ABWRs can provide us. The ABWR has already been awarded a U.S. standard license for having proved that it can comply with the requirements of international regulatory systems with an ample margin. There are also many construction programs with ABWRs progressing both domestically and abroad, suggesting that it has won recognition as an international standard plant. We will do our utmost to perfect the operation and maintenance records of Kashiwazaki-Kariwa Units No.6 and 7, which is the top runner among ABWRs, and to make known the superiority of this reactor to the world. (J.P.N.)

The research committee of Chuetsu-oki earthquake influences to Kashiwazaki-Kariwa Nuclear Power Station. How to press-release and take care of expression in articles at press side

International Nuclear Information System (INIS)

Hamamoto, Kazuko; Narabayashi, Tadashi; Kobayashi, Masahide; Akizuki, Teruo; Onishi, Hidetoshi

2009-01-01

As for the influences of Chuetsu-Oki Earthquake on Kashiwazaki-Kariwa Nuclear Power Station, we can conclude that 'the safety function of the nuclear power station, that is, 'Shutdown', 'Cooling down' and 'Isolating' functioned as per designed even against an earthquake beyond assumption, and fundamental nuclear safety could assured'. Nevertheless, it is said that one of the causes that harmful rumor had spread was due to mass communication media. In the press reports on some nuclear power station when affected by an earthquake and on trouble in the nuclear power station in the future, we will propose that the publication should be really useful for habitants and citizens and be promoted in the positive and expected direction, in order not to make the same mistake as this time. JSME should aim at implement the above-mentioned proposal under cooperation with other academic societies and organizations. (author)

The KASKA project - a Japanese medium-baseline reactor-neutrino oscillation experiment to measure the mixing angle $\\theta_{13}$ -

OpenAIRE

Kuze, Masahiro; Collaboration, for the KASKA

2005-01-01

A new reactor-neutrino oscillation experiment, KASKA, is proposed to measure the unknown neutrino-mixing angle $\\theta_{13}$ using the world's most powerful Kashiwazaki-Kariwa nuclear power station. It will measure a very small deficit of reactor-neutrino flux using three identical detectors, two placed just close to the sources and one at a distance of about 1.8km. Its conceptual design and physics reach are discussed.

Follow-up IAEA mission in relation to the findings and lessons learned from the 16 July 2007 earthquake at Kashiwazaki-Kariwa NPP - 'The Niigataken Chuetsu-oki earthquake', Tokyo and Kashiwazaki-Kariwa NPP, Japan, 28 January - 1 February 2008. Mission report. V. 1

International Nuclear Information System (INIS)

2008-01-01

On 16 July 2007, a strong earthquake, the Niigataken Chuetsu-oki earthquake, with a moment magnitude of 6.6 (M JMA =6.8 according to the Japanese Meteorological Agency), occurred at 10:13 h local time with its hypocentre below the seabed of the Jo-chuetsu area in Niigata prefecture (37 deg. 33' N, 138 deg. 37'E) in Japan, affecting the Kashiwazaki-Kariwa Nuclear Power Plant (NPP) located approximately 16 km south of its epicentre. Kashiwazaki-Kariwa NPP is the biggest nuclear power plant site in the world. It is located in the Niigata prefecture, in the northwest coast of Japan, and it is operated by Tokyo Electric Power Company (TEPCO). The site has seven units with a total of 7965 MW net installed capacity. Five reactors are of BWR type and two reactors are of ABWR type. The five BWR units entered commercial operation between 1985 and 1994 and the two ABWRs in 1996 and 1997, respectively. Following this event, the Government of Japan through the Nuclear and Industrial Safety Agency (NISA) requested the IAEA to carry out a fact finding mission with the main purpose of identifying the preliminary findings and lessons learned from this event in order to share them with the international nuclear community. This first mission took place from 6 - 10 August 2007 and the mission report of the August 2007 mission is available on the IAEA web page http://www.iaea.org. The purpose of the second IAEA mission was to conduct - six months after the event - a follow-up of the preliminary findings of the August 2007 mission on the basis of the results available in January 2008 of the related studies and investigations performed. In accordance with the terms of reference for the follow-up mission and the availability of results from the performed studies and investigations, the scope of the follow-up mission focussed on three subject areas: (1) seismic design basis - design basis ground motions, including the evaluation of the seismic hazard ; (2) plant behaviour - integrity

Reactor Neutrino Oscillations: KamLAND and KASKA

International Nuclear Information System (INIS)

Suekane, F.

2006-01-01

Nuclear reactors generate a huge number of low energy ν-bar e 's. The reactor neutrinos have been used to study properties of neutrinos since its discovery a half century ago. Recently, KamLAND group finally discovered reactor neutrino oscillation with average baseline 180 km. According to the 3 flavor scheme of standard theory and measured oscillation parameters so far, the reactor neutrino is expected to perform another type of small oscillation at a baseline 1.8 km. KASKA experiment is a project to detect this small oscillation and to measure the last neutrino mixing angle θ 13 by using the world most powerful reactor complex, Kashiwazaki-Kariwa nuclear power station. In this proceedings, phenomena of neutrino oscillation and the two reactor oscillation experiments, KamLAND and KASKA, are introduced

On-site underground background measurements for the KASKA reactor-neutrino experiment

International Nuclear Information System (INIS)

Furuta, H.; Sakuma, K.; Aoki, M.; Fukuda, Y.; Funaki, Y.; Hara, T.; Haruna, T.; Ishihara, N.; Katsumata, M.; Kawasaki, T.; Kuze, M.; Maeda, J.; Matsubara, T.; Matsumoto, T.; Miyata, H.; Nagasaka, Y.; Nakagawa, T.; Nakajima, N.; Nitta, K.; Sakai, K.; Sakamoto, Y.; Suekane, F.; Sumiyoshi, T.; Tabata, H.; Tamura, N.; Tsuchiya, Y.

2006-01-01

On-site underground background measurements were performed for the planned reactor-neutrino oscillation experiment KASKA at Kashiwazaki-Kariwa nuclear power station in Niigata, Japan. A small-diameter boring hole was excavated down to 70m underground level, and a detector unit for γ-ray and cosmic-muon measurements was placed at various depths to take data. The data were analyzed to obtain abundance of natural radioactive elements in the surrounding soil and rates of cosmic muons that penetrate the overburden. The results will be reflected in the design of the KASKA experiment

Next generation light water reactors

International Nuclear Information System (INIS)

Omoto, Akira

1992-01-01

In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

Geoengineering and seismological aspects of the Niigata-Ken Chuetsu-Oki earthquake of 16 July 2007

Science.gov (United States)

Kayen, R.; Brandenberg, S.J.; CoIlins, B.D.; Dickenson, S.; Ashford, S.; Kawamata, Y.; Tanaka, Y.; Koumoto, H.; Abrahamson, N.; Cluff, L.; Tokimatsu, K.

2009-01-01

The M6.6 Niigata-Ken Chuetsu-Oki earthquake of 16 July 2007 occurred off the west coast of Japan with a focal depth of 10 km, immediately west of Kashiwazaki City and Kariwa Village in southern Niigata Prefecture. Peak horizontal ground accelerations of 0.68 g were measured in Kashiwazaki City, as well as at the reactor floor level of the world's largest nuclear reactor, located on the coast at Kariwa Village. Liquefaction of historic and modern river deposits, aeolian dune sand, and manmade fill was widespread in the coastal region nearest the epicenter and caused ground deformations that damaged bridges, embankments, roadways, buildings, ports, railways and utilities. Landslides along the coast of southern Niigata Prefecture and in mountainous regions inland of Kashiwazaki were also widespread affecting transportation infrastructure. Liquefaction and a landslide also damaged the nuclear power plant sites. This paper, along with a companion digital map database available at http://walrus.wr.usgs.gOv/infobank/n/nii07jp/html/n-ii-07-jp.sites.kmz, describes the seismological and geo-engineering aspects of the event. ?? 2009, Earthquake Engineering Research Institute.

Nuclear reactor system for ABWR

International Nuclear Information System (INIS)

Miyano, Hiroshi; Kitagawa, Koji

1997-01-01

Various tests and measurements were performed during the pre-operational test run of Unit No. 6 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station, the first advanced boiling water reactor (ABWR) unit in the world, and the design and performance adequacy of the ABWR were confirmed. The realization of the ABWR in Japan took about 20 years. It was decided that technologies for the reactor internal pump (RIP) and the fine-motion control rod drive (FMCRD), which had been applied in Europe, would be incorporated in the ABWR aiming at simplification of its structure and operation. These main components were evaluated, modified and verified in consideration of the unique Japanese environment, such as seismic conditions, through a joint study program with Japanese utilities as well as an improvement and standardization program in cooperation with the government. In addition to incorporating RIP and FMCRD technologies, the ABWR also has improved features in terms of the design of the reactor pressure vessel and internals, as well as automated servicing equipment for the RIP, FMCRD, and primary containment vessel. (author)

Estimation of strong motions on free rock surface. Identification of soil structures and strong motions on free rock surface in Kashiwazaki-Kariwa nuclear power plant during the 2007 Niigataken Chuetsu-oki earthquake

International Nuclear Information System (INIS)

Saguchi, Koichiro; Masaki, Kazuaki; Irikura, Kojiro

2009-01-01

Very strong ground motions (maximum acceleration 993 cm/s 2 in the borehole seismometer point of -255m in depth) were observed in the Kashiwazaki Kariwa Nuclear Power Plant during the Niigataken Chuetsu-oki Earthquake on July 16, 2007. In this study, we tried to develop new method, which can simulate waveforms on free rock surface by using the bore hole records. We identified the underground structure model at the Service Hall from aftershock records observed in vertical array, using the simulated annealing method (Ingber(1989)). Based on numerical experiments it is identified that S-wave velocity and Q values of individual layers are inverted very well. Strong motion records of main shock observed by the bore hole seismometers were simulated by using one-dimensional multiple reflection method. In this study, non-linear effect is considered by introducing non-linear coefficient c(f) for under coming wave from surface. The maximum acceleration and phase characteristics in simulated waveforms are similar to the observed one. It means that our method is useful for simulate strong motion in non-linear region. Finally, strong motions on the free rock surface at the Service Hall during the main shock are simulated. The maximum acceleration of EW component on free rock surface is estimated to be 1,207 cm/s 2 . (author)

Seismic Observation in Deep Boreholes and Its Applications - Workshop Proceedings, Niigata Institute of Technology, Kashiwazaki, Japan

International Nuclear Information System (INIS)

2014-01-01

The Kashiwazaki WS was held to develop the recommendations made at the Tsukuba WS entitled 'Seismic Input Motions Incorporating Recent Geological Studies' which was held in November 2004 in Tsukuba City in Japan (hereinafter, the 'Tsukuba WS'). At the Tsukuba WS, the state of the art in defining realistic seismic input for the design and re-evaluation of nuclear facilities as well as advances in seismic source characterization of fault zones using data from deep geological investigations and their possible contribution to improving seismic input definitions were reviewed. Further, the importance and necessity of cooperation between seismology and geology in order to decrease uncertainty in seismic input definition were emphasized. After the Tsukuba WS, the Niigata-ken Chuetsu-oki Earthquake (NCOE, M=6.8) occurred near the Kashiwazaki-Kariwa NPP site. In this earthquake, a focusing effect of ground motion was observed at the Kashiwazaki-Kariwa NPP site, which caused locally amplified ground motion. Units 1 and 2 showed significantly higher responses of more than 50 percent compared to Units 3 to 7. This was thought to be caused by the fact that seismic waves were focused on Units 1 and 2 due to the irregular structure under the site. A similar effect was observed at the Hamaoka site in the earthquake that occurred in Suruga Bay in 2009. Only Unit 5 showed a double or larger response to this earthquake, and similar phenomena were found only for events with hypocenter in the narrow direction from the site. This was also thought to be caused by the irregular geological structure under the site. In the 2011 off the Pacific Coast of Tohoku Earthquake, the peak ground accelerations (PGA) of the Dai-ichi NPP site were about twice as large as those of the Dai-ni NPP site, although the distance between these sites is only slightly more than 10 km. Further, large differences in the PGAs were found among each of the units of the Dai-ichi NPP site. For example, the PGA of Unit

On the effect of the 3-D regional geology on the seismic design of critical structures: the case of the Kashiwazaki-Kariwa Nuclear Power Plant

Science.gov (United States)

Gatti, F.; Lopez-Caballero, F.; Clouteau, D.; Paolucci, R.

2018-05-01

In this study, numerical investigation is performed on a realistic source-to-site earthquake scenario, with the aim to assess the role of complex 3-D geological structures on the predicted wavefield. With this respect, the paper pointedly targets the seismic response of nuclear power plants in near-field conditions and the verification of some simplified assumptions commonly adopted for earthquake ground motion prediction and site effects analysis. To this purpose, the Kashiwazaki-Kariwa Nuclear Power Plant (Japan) is assumed as reference case-study. In 2007, the nuclear site and its surroundings were struck by the Niigata-Ken Chūetsu-Oki seismic sequence, which caused some of the peak ground motion design limits to be largely overpassed. The dense observation network deployed at the site recorded a highly incoherent and impulsive earthquake ground motion. Many studies argued that the intricate syncline-anticline geology lying underneath the nuclear facility was highly responsible of the observed seismic response. Therefore, a physics-based numerical model of the epicentral area is built-up (≈60 km wide) and tested for small aftershocks, so to discount the effect of extended source on the synthetic site-response. The numerical model (based on the Spectral Element Method) reproduces the source-to-site wave propagation by embracing the effects of the surface topography along with the presence of the Japan Sea (i.e. the bathymetry, the coastline and the fluid-solid interaction). Broad-band (0-5 Hz) synthetic waveforms are obtained for two different aftershocks, located at the two opposite sides of the nuclear facility, aiming to assess the influence of the incidence angle the radiated wave field impinges the foldings beneath it. The effect of the folding presence is assessed by comparing it to a subhorizontally layered geology, in terms of numerical outcome, and by highlighting the differences with respect to the observations. The presence of an intricate geology

Development of artificial soft rock

International Nuclear Information System (INIS)

Kishi, Kiyoshi

1995-01-01

When foundation base rocks are deeper than the level of installing structures or there exist weathered rocks and crushed rocks in a part of base rocks, often sound artificial base rocks are made by substituting the part with concrete. But in the construction of Kashiwazaki Kariwa Nuclear Power Station of Tokyo Electric Power Co., Inc., the foundation base rocks consist of mudstone, and the stiffness of concrete is large as compared with the surrounding base rocks. As the quality of the substituting material, the nearly same stiffness as that of the surrounding soft rocks and long term stability are suitable, and the excellent workability and economical efficiency are required, therefore, artificial soft rocks were developed. As the substituting material, the soil mortar that can obtain the physical property values in stable form, which are similar to those of Nishiyama mudstone, was selected. The mechanism of its hardening and the long term stability, and the manufacturing plant are reported. As for its application to the base rocks of Kashiwazaki Kariwa Nuclear Power Station, the verification test at the site and the application to the base rocks for No. 7 plant reactor building and other places are described. (K.I.)

Expectations of stake holders voice from the location

International Nuclear Information System (INIS)

Shinada, H.

2007-01-01

In this presentation, the expectations for transparency of the nuclear regulatory authorities are discussed from the viewpoint of a local government, where the nuclear power station is sited. Since the problems yet to be solved by the local government, are visibly existent right in front of our eyes, both the policy information and the technical information required for that purpose must he altogether trustworthy. Consequently, established transparency of the regulatory authorities is also an extremely important point for the local government. To begin with, the outline of Kariwa-mura, Kariwa-gun, Niigata Prefecture, where I myself serve as village mayor, the overview of the Kashiwazaki-Kariwa Nuclear Power Station and the relationship between Kariwa-mura and the power station are explained. Next in place, the peculiarity of Kariwa-mura involved in the nuclear power station will be mentioned after defining the transparency required in relation to the regulations in order to clarify in what respect we are seeking accountability from the regulatory authorities. In conclusion, the roles of the local residents (local government) involved in the nuclear safety regulations are discussed introducing the 'Regional Association for Securing the Transparency of Kashiwazaki Nuclear Power Station' as an sure example to the solution for securing transparency. (author)

Support system development for evacuation plan decision in nuclear plant disaster

International Nuclear Information System (INIS)

Fujita, Masahiko; Takayama, Jun-ichi; Nakayama, Sho-ichiro; Ushiba, Takashi

2011-01-01

These days, our interest in nuclear plant accidents has increased, and civic actions for them have also been activated. Therefore, improvement of the disaster prevention planning to nuclear plant accidents is requested. In this study, we developed a microscopic traffic simulation system for evacuation plan near the nuclear plant as a system which supports to examine the disaster prevention planning, and applied the system to Kashiwazaki-Kariwa nuclear plant area. Furthermore, the risk of each region near the nuclear plant disaster from the viewpoint of wind direction and the population was considered, the importance of each evacuation simulation was examined. As a result, we found that the present plan Kashiwazaki-Kariwa made has the problem on evacuation routes and others. (author)

Approach of seismic upgrading in Kashiwazaki-Kariwa Nuclear Power Plant

International Nuclear Information System (INIS)

Sato, Hitoshi

2009-01-01

Because guide for reviewing seismic design of nuclear power reactor facilities was reworked in 2006, we formulated new Design Base Seismic Motion Ss, and we are doing evaluation of seismic safety (back-check). In Japan, depending on aseismatic importance, equipments are classified into S-class, B-class and C-class. For S-class equipments, we evaluate it on the basis of new Ss, and do seismic upgrading. For B-class and C-class equipments, we do seismic upgrading voluntarily on the basis of the experiences of the Niigataken Chuetsu-Oki (NCO) Earthquake. (author)

Construction of the advanced boiling water reactor in Japan

International Nuclear Information System (INIS)

Natsume, Nobuo; Noda, Hiroshi

1996-01-01

The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7

Development, operating experience and future plan of ABWR in Japan

International Nuclear Information System (INIS)

Ujihara, N.

2001-01-01

From 1974 to 1978 nuclear power plants' capacity factor decreased to a low level at the Tokyo Electric Power Company (TEPCO) for stress corrosion cracking of stainless steal pipe, thermal fatigue crack of feed water spurger and vibration trouble of LPRM, etc. So, for improving that situation, we gathered the most useful BWR technology in the world at that time and we started the ABWR project to develop an ideal BWR. The goals of this development were: (1) Enhanced safety and reliability; (2) Reduced occupational radiation exposure and radioactive waste; (3) Enhanced operability and maneuverability; and (4) Improved economy. Significant improvements were induced by the adoption of the reactor internal pump, fine motion control rod drive mechanism, integral type reinforced concrete containment vessel and digital control system. After about 10 years R and D, we succeeded in the development of the compact and economic plant, ABWR (1356MW(e)) whose building volume had been reduced to 76%, compared with the previous BWR (1100MW(e)). The construction of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6, which is the first ABWR plant in the world, was started September 1991, and that of unit No. 7 was started in February 1992. We used effective construction methods, such as large block construction method and all weather construction method, to shorten the construction duration from first concrete to commercial operation start to 51.5 months. The commercial operation start of Kashiwazaki-Kariwa Nuclear Power Plant unit No. 6 was November 1996 and that of unit No. 7 was July 1997. After that, unit No. 6 had three times the annual outage and unit No. 7 had two times the annual outage. Both plants have good operating experiences. From commercial operation start to end of August 2000, the capacity factor of unit No. 6 is 86.4% and that of unit No. 7 is 86.9%. Now, in Taiwan, ABWR plants are under construction at Lungmen unit No. 1 and 2. In Japan, ABWR plants are under

Consideration for standard earthquake vibration (1). The Niigataken Chuetsu-oki Earthquake in 2007

International Nuclear Information System (INIS)

Ishibashi, Katsuhiko

2007-01-01

Outline of new guideline of quakeproof design standard of nuclear power plant and the standard earthquake vibration are explained. The improvement points of new guideline are discussed on the basis of Kashiwazaki-Kariwa Nuclear Power Plant incidents. The fundamental limits of new guideline are pointed. Placement of the quakeproof design standard of nuclear power plant, JEAG4601 of Japan Electric Association, new guideline, standard earthquake vibration of new guideline, the Niigataken Chuetsu-oki Earthquake in 2007 and damage of Kashiwazaki-Kariwa Nuclear Power Plant are discussed. The safety criteria of safety review system, organization, standard and guideline should be improved on the basis of this earthquake and nuclear plant accident. The general knowledge, 'a nuclear power plant is not constructed in the area expected large earthquake', has to be realized. Preconditions of all nuclear power plants should not cause damage to anything. (S.Y.)

Operating data monitoring and fatigue evaluation systems and findings for boiling water reactors in Japan

International Nuclear Information System (INIS)

Maekawa, O.; Kanazawa, Y.; Takahashi, Y.; Tani, M.

1995-01-01

The reactor pressure vessel (RPV) is one of the most critical components of a boiling water reactor (BWR) when utilities think about plant life extension (PLEX). Design stress analysis sometimes reports very high fatigue usage factors for such portions of RPVs as stud bolt, feedwater nozzle and support skirt.In order to evaluate design margin and to eliminate excessive conservatism in this design analysis to pave the way for PLEX, Japanese BWR utilities jointly with BWR manufacturers in Japan established a programme (1) to acquire plant operational data on line for specific parameters used in stress analysis, (2) to evaluate margin in the design using measured plant data best estimate boundary conditions for stress analysis, and (3) to establish a simplified fatigue analysis method for BWR RPV.A plant data acquisition system, named OPEDAS, has been developed and installed in Tokyo Electric Power Company's 1100MWe BWR at its Kashiwazaki-Kariwa nuclear power plant. Best estimate stress analysis using measured in-plant data has been carried out and the results show considerable margin in fatigue usage factor over the design. A simplified fatigue analysis method using in-plant data has been developed with the Green's function, although some limitations have been identified for its use. ((orig.))

Start of the construction of initial advanced BWRs

International Nuclear Information System (INIS)

Yamada, Yasuo; Furukawa, Hideyasu; Yoshida, Tomiharu

1992-01-01

The construction of first advanced BWRs (ABWR), No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc., was started, and the commercial operation is scheduled to begin in 1996 and 1997. The ABWR with the unit capacity of 1356 MW which is the largest class in the world has been developed aiming at the further improvement of safety, operation performance, economical efficiency and so on. Hitachi Limited carried out the detailed review on new parts design and expected the perfection so as to make the first machine into the plant maintaining high reliability. The course of development of the ABWR, the outline of No. 6 and No. 7 plants in Kashiwazaki Kariwa Nuclear Power Station, the review of the design of new parts, the establishment of the technology of manufacturing RCCVs, the verification of the maintenance and checkup of internal pumps and improved type control rod driving mechanism and so on are reported. The ABWR is Japanese third improved type standard plant, but also it is expected to be adopted worldwide. (K.I.)

Acid Rain and Snow in Kashiwazaki City.

OpenAIRE

小野寺, 正幸; 富永, 禎秀; 竹園, 恵; 大金, 一二; Onodera, Masayuki; Tominaga, Yoshihide; Takesono, Satoshi; Oogane, Katsuji

2002-01-01

This paper described the actual condition of acid rain and snow and their influence of a winter monsoon in Kashiwazaki city. For 7 months from September in 2001 to March in 2002, the pH value was measured in rain or snow. The minimum of pH value observed was 3.9 for the 7 months. The day which observed pH

Safety aspect of digital reactor protection system in Japan

International Nuclear Information System (INIS)

Ogiso, Zen-Ichi

1998-01-01

It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

Evaluation of effectiveness of monorail type inspection robot at nuclear power plant

International Nuclear Information System (INIS)

Nakagawa, Katsuei; Kimura, Motohiko; Ito, Takao; Sasaki, Keiichi.

1991-01-01

An inspection robot, with a TV camera, infrared camera and microphone as sensors, was tentatively installed in the main steam tunnel room of Kashiwazaki-Kariwa N.P.S. no.1 to evaluate its effectiveness in actual plant use. After a one and a half-year run, it appeared to have the ability to perform daily patrol tasks inplace of human beings. (author)

Design concepts for an integrated control room used as a site-wide operations facility

International Nuclear Information System (INIS)

Simon, B.H.; Raghavan, R.; Ujita, H.; Utena, S.; Sakuma, A.; Itoh, T.; Fukura, M.; Ono, I.

1995-01-01

The concept of an Integrated Main Control Room (IMCR) evolved from surveys conducted by Tokyo Electric Power Company (TEPCO) with plant managers and workers as their existing GE-type boiling water reactors (BWRs) on the need for improved operating conditions in a new generation of reactors being developed for the next century (around 2010). These reactors will be a further enhancement of the advanced boiling water reactors (ABWRs) now being constructed at the Kashiwazaki-Kariwa site in Japan (no.6 and no.7). TEPCO also saw a need for new thinking on control room design because of projected social conditions in Japan for the 21st century. These projections forecast a smaller number of skilled engineering graduates and those graduates less willing to work in nuclear power because such work is seen as unappealing, conducted in remote geographical locations, and requiring extensive night duty. As one solution to reducing operator burden and decreasing the night shift staff, while making nuclear plant operation more interesting for the operators and reducing labor and construction costs, the IMCR was conceived. (author)

Development of the advanced on-line BWR core monitoring system TiARA

International Nuclear Information System (INIS)

Kobayashi, Yoko; Yamazaki, Hiroshi

1996-01-01

Development of an integrated computer environment to support plant operators and station nuclear engineers is a recent activity. In achieving this goal, an advanced on-line boiling water reactor (BWR) core monitoring system: TiARA has been developed by Toden Software. An integrated design approach was performed through the introduction of recent computer technologies, a sophisticated human/machine interface (HMI) and an advanced nodal method. The first prototype of TiARA was ready in early 1996. This prototype is now undergoing a field test at Kashiwazaki-Kariwa unit 6. After successful completion of this test, the authors will have achieved the following goals: (1) consistency between on-line core monitoring system and off-line core management system; (2) an enhanced HMI and database; (3) user-friendly operability and maintainability; (4) system development from the utilities' standpoint to fully satisfy operator needs

Working material. IAEA seismic safety of nuclear power plants. International workshop on lessons learned from strong earthquake

International Nuclear Information System (INIS)

2008-08-01

The International Workshop on Lessons Learned from Strong Earthquake was held at Kashiwazaki civic plaza, Kashiwazaki, Niigata-prefecture, Japan, for three days in June 2008. Kashiwazaki-Kariwa NPP (KK-NPP) is located in the city of Kashiwazaki and the village of Kariwa, and owned and operated by Tokyo Electric Power Company Ltd. (TEPCO). After it experienced the Niigata-ken Chuetsu-oki earthquake in July 2007, IAEA dispatched experts' missions twice and held technical discussions with TEPCO. Through such activities, the IAEA secretariat and experts obtained up-dated information of plant integrity, geological and seismological evaluation and developments of the consultation in the regulatory framework of Japan. Some of the information has been shared with the member states through the reports on findings and lessons learned from the missions to Japan. The international workshop was held to discuss and share the information of lessons learned from strong earthquakes in member states' nuclear installations. It provided the opportunity for participants from abroad to share the information of the recent earthquake and experience in Japan and to visit KK-NPP. And for experts in Japan, the workshop provided the opportunity to share the international approach on seismic-safety-related measures and experiences. The workshop was organised by the IAEA as a part of an extra budgetary project, in cooperation with OECD/NEA, hosted by Japanese organisations including Nuclear and Industrial Safety Agency (NISA), Nuclear Safety Commission (NSC), and Japan Nuclear Energy Safety Organization (JNES). The number of the workshop participants was 70 experts from outside Japan, 27 countries and 2 international organisations, 154 Japanese experts and 81 audience and media personnel, totalling to 305 participants. The three-day workshop was open to the media including the site visit, and covered by NHK (the nation's public broadcasting corporation) and nation-wide and local television

Seismic-resistant design of nuclear power stations in Japan, earthquake country. Lessons learned from Chuetsu-oki earthquake

International Nuclear Information System (INIS)

Irikura, Kojiro

2008-01-01

The new assessment (back-check) of earthquake-proof safety was being conducted at Kashiwazaki-Kariwa Nuclear Power Plants, Tokyo Electric Co. in response to a request based on the guideline for reactor evaluation for seismic-resistant design code, revised in 2006, when the 2007 Chuetsu-oki Earthquake occurred and brought about an unexpectedly huge tremor in this area, although the magnitude of the earthquake was only 6.8 but the intensity of earthquake motion exceeded 2.5-fold more than supposed. This paper introduces how and why the guideline for seismic-resistant design of nuclear facilities was revised in 2006, the outline of the Chuetsu-oki Earthquake, and preliminary findings and lessons learned from the Earthquake. The paper specifically discusses on (1) how we may specify in advance geologic active faults as has been overlooked this time, (2) how we can make adequate models for seismic origin from which we can extract its characteristics, and (3) how the estimation of strong ground motion simulation may be possible for ground vibration level of a possibly overlooked fault. (S. Ohno)

Problems with geologic fault assessment in the safety analysis of earthquake-resistance for nuclear power plants

International Nuclear Information System (INIS)

Suzuki, Yasuhiro; Nakata, Takashi; Watanabe, Mitsuhisa

2008-01-01

The Niigata-ken Chuetsu-oki Earthquake, occurring on July 16, 2007 in Niigata-prefecture, Japan, the magnitude of which was only 6.8, brought about an unexpectedly huge tremor in the area of Kashiwazaki-Kariwa nuclear plants, Tokyo Electric Power Co. The present paper points out that the huge tremor they experienced could be within the scope of expectation if an adequate analysis was performed for the investigation results on acoustic seafloor exploration done during the 1980's and presented at application for go-ahead for construction of the nuclear reactor and also at the occasion of official safety analysis. The present examination demonstrates clearly the existence of active faults on a large scale in the hypocentral region of the earthquake in the offshore seabed. This fact might be of extremely serious nature and the authors recommend to learn what may be a cause of overlooking the existing fault. (S. Ohno)

TEPCO plans to construct Higashidori Nuclear Power Station

International Nuclear Information System (INIS)

Tsuruta, Atsushi

2008-01-01

In 2006, TEPCO submitted to the government plans for the construction of Higashidori Nuclear Power Station. The application was filed 41 years after the project approved by the Higashidori Village Assembly. This nuclear power station will be the first new nuclear power plant constructed by TEPCO since the construction of Units No.6 and 7 at the Kashiwazaki Kariwa Nuclear Power Station 18 years ago. Higashidori Nuclear Power Station is to be constructed at a completely new site, which will become the fourth TEPCO nuclear power station. Higashidori Nuclear Power Station Unit No.1 will be TEPCO's 18th nuclear reactor. Unit No.1 will be an advanced boiling water reactor (ABWR), a reactor-type with a proven track record. It will be TEPCO's third ABWR. Alongside incorporating the latest technology, in Higashidori Nuclear Power Station Unit No.1, the most important requirement is for TEPCO to reflect in the new unit information and experience acquired from the operation of other reactors (information and experience acquired through the experience of operating TEPCO's 17 units at Fukushima Daiichi Nuclear Power Station, Fukushima Daini Nuclear Power Station and Kashiwazaki Kashiwa Nuclear Power Station in addition to information on non-conformities at nuclear power stations in Japan and around the world). Higashidori Nuclear Power Station is located in Higashidori-Village (Aomori Prefecture) and the selected site includes a rich natural environment. From an environmental perspective, we will implement the construction with due consideration for the land and sea environment, aiming to ensure that the plant can co-exist with its natural surroundings. The construction plans are currently being reviewed by the Nuclear and Industrial Safety Agency. We are committed to making progress in the project for the start of construction and subsequent commercial operation. (author)

Earthquake in Japan: The IAEA mission gives its report

International Nuclear Information System (INIS)

Anon.

2007-01-01

Following the seism that occurred on the 16. july 2007 in Japan (magnitude 6.6 on Richter scale), an IAEA mission has inspected the nuclear power plant of Kashiwazaki Kariwa at the beginning of August. The mission has estimated that the safety of the installation has been provided during and after the earthquake, in spite of the fact that the earthquake has gone past the seism level taken as reference in the conception of the nuclear facility. The systems and the components were in a better state that it could be imagined after a such earthquake. The release have been under the authorised thresholds. At the moment of the seism, three reactors were running on the seven ones of the nuclear power plant, and stopped automatically. The unit 2 that started up, has also stopped automatically. The reactors 1, 5 and 6 were stopped for maintenance. Water poured out coming from the spent fuel storage pool because of the earth tremors. It was picked and thrown out by the release pipe to the sea without notable impact on environment ( volume 1.2 m 3 ). One hundred of containers was overturned. Traces of iodine, chromium 51 and cobalt 60 have been found in the ventilation filters ( reactor 7) these elements have been released in atmosphere in very low quantities. (N.C.)

ABWR construction experience in Japan

International Nuclear Information System (INIS)

Kawahata, Junichi; Miura, Jun; Saito, Fuyuki; Mori, Hiroya

2003-01-01

The construction of Kashiwazaki-Kariwa Nuclear Power Station Unit No. 6 and 7 (K6 and 7) owned by Tokyo Electric Power Company (TEPCO), have been in service since November 1996 and September 1997 respectively, and are the first ABWRs in the world. Hitachi, Ltd. and Toshiba Corporation took part in the projects and achieved safety construction in 52- month schedule from rock inspection to turnover in cooperation with TEPCO and other contractors with broad experience in BWRs. Taking advantage of experience of the Kashiwazaki construction, the ABWR construction methods realize to satisfy with requests from electric power companies and to fit site conditions. The Chubu Electric Power Company's Hamaoka Nuclear Power Station Unit No. 5 (Hamaoka 5) and Hokuriku Electric Power Company's Shika Nuclear Power Station Unit No. 2 (Shika 2), which are ABWRs, completed rock inspection in May 2000 and June 2001 respectively, and are under construction. The ABWR construction methods are proved their validity through four actual construction projects. Safe constructions with the highest quality have been advanced thus further aiming at the introduction of the latest technologies and contributing to reduce plant construction costs and also to shorten schedule. (author)

Man-machine interface systems and operator training program for ABWR in Japan

International Nuclear Information System (INIS)

Kunito, Susumu

2004-01-01

The Tokyo Electric Power Company (TEPCO) has developed a new Main Control Room design for the Advanced Boiling Water Reactor (ABWR) to improve man-machine interface. New configuration of panels and enhanced automation are some of the features of the ABWR type Main Control Room design. Various technologies such as Cathode Ray Tubes (CRTs) and Flat Displays (FDs) with touch-sensitive operations are contributed to the development of the ABWR type control room design. This design will be first applied to Kashiwazaki-Kariwa Nuclear Power Station unit 6 (K-6). To train the operators sufficiently, TEPCO reviewed the operator training program. Compared with the conventional training, new training menu will be added and the training of ABWR operators will be started 6 months earlier. An ABWR simulator is under construction and training using this simulator is scheduled to be started in August 1994, which is 18 months before fuel loading of K-6. We are reviewing malfunction modes on the simulator. (author)

Evaluation of near-surface attenuation of S-waves based on PS logging and vertical array seismic observation

International Nuclear Information System (INIS)

Kobayashi, Genyu

2014-01-01

As a result of the lessons learned from the experience of Kashiwazaki-Kariwa NPP due to the 2007 Niigata Chuetsu Oki Earthquake, it has become clear that a rational method of near-surface attenuation characteristics covering a depth range from engineering bedrock to seismic bedrock urgently needs to be established. JNES performed PS logging and vertical array seismic ground motion observation at a soil ground site (SODB 1. site), sedimentary rock site, and an igneous rock site (SODB 2. site), and proposed an evaluation method of attenuation characteristics (site characteristics) for the deep underground. (author)

ABWR (K-6/7) construction experience (computer-based safety system)

International Nuclear Information System (INIS)

Yokomura, T.

1998-01-01

TEPCO applied a digital safety system to Kashiwazaki-Kariwa Nuclear Power Station Unit Nos. 6 and 7, the world's first ABWR plant. Although this was the first time to apply a digital safety logic system in Japan, we were able to complete construction of K-6/7 very successfully and without any delay. TEPCO took a approach of developing a substantial amount of experience in digital non- safety systems before undertaking the design of the safety protection system. This paper describes the history, techniques and experience behind achieving a highly reliable digital safety system. (author)

Recent technology for BWR nuclear steam turbine unit

International Nuclear Information System (INIS)

Moriya, Shin-ichi; Masuda, Toyohiko; Kashiwabara, Katsuto; Oshima, Yoshikuni

1990-01-01

As to the ABWR plants which is the third improvement standard boiling water reactor type plants, already the construction of a plant of 1356 MWe class for 50 Hz is planned. Hitachi Ltd. has accumulated the technology for the home manufacture of a whole ABWR plant including a turbine. As the results, the application of a butterfly type combination intermediate valve to No.5 plant in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc., which began the commercial operation recently and later plants, the application of a moisture separating heater to No.4 plant in Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., which is manufactured at present and later plants and so on were carried out. As to the steam turbine facilities for nuclear power generation manufactured by Hitachi Ltd., three turbines of 1100 MWe class for 50 Hz and one turbine for 60 Hz are in operation. As the new technologies for the steam turbines, the development of 52 in long last stage blades, the new design techniques for the rotor system, the moisture separating heater, the butterfly type combination intermediate valve, cross-around pipes and condensate and feedwater system are reported. (K.I.)

The research committee of Chuetsu-oki earthquake influences to Kashiwazaki-Kariwa Nuclear Power Station. Importance classification

International Nuclear Information System (INIS)

Mizumachi, Wataru; Kobayashi, Masahide

2009-01-01

Conventionally, the design of a nuclear reactor has been performed from a viewpoint of a safety function and the importance on earthquake-proof on the basis of not giving off the mainly included radioactivity outside. In this Niigataken-Chuetsuoki earthquake, there is almost no damage to the system, components and structure on safe also in the earthquake beyond assumption, and the validity of the design was checked. But, the situation peculiar to a big earthquake was also generated. The emergency plan room which should serve as a connection center with the exterior was not able to open a door and use at the beginning. Fire-extinguishing system piping fractured and self-defense fire fighting was not made. And so on. Discussion from the following three viewpoints was performed. (1) The importance from a viewpoint which should maintain a function also with the disaster in case of an earthquake like an emergency plan room etc. (2) In the earthquake, since the safe system and un-safe system was influenced, the importance from a viewpoint which may have influence safely inquired when the un-safe system broke down. (2) Although it was not directly related safely, discussion from a viewpoint which influences fear of insecurity, such as taking out smoke, for example, was performed. (author)

News from the world

International Nuclear Information System (INIS)

Anon.

2008-01-01

This document gathers short notices concerning the nuclear industry. Among them we find the following ones. A nuclear power plant will be built in Bulgaria by a Russian company, the plant is due to operate in 2013. An IAEA expert team has visited the Kashiwazaki-Kariwa site, a Japanese nuclear power plant involving 7 reactors and that underwent a severe earthquake in July 2007 (6.8 grades on the Richter scale). It appears that no major damage occurred on the reactors and that the various infra-structures involved in nuclear safety behaved better than expected. Kazakhstan has increased its uranium production by 25% to reach 6637 tonnes of extracted uranium. The Konrad well situated on the site of the ancient iron mines of Salzgitter, has been agreed by German authorities for the storage of radioactive wastes of low or intermediate level. The equipping works will last 4 years and the storage will begin in 2013. An opinion poll made in Sweden shows that almost half of the population support the construction of new nuclear power plants while Sweden is due to decommission all of its reactors within 20 years. The Russian authorities have announced the creation of a university dedicated to nuclear engineering. India has joined the international cooperation around the Jules Horowitz reactor (RJH reactor). China has become a full-fledged member of the Generation IV forum. American authorities have given their green light for a cooperation with Turkey on civil nuclear energy. (A.C.)

The development of ABWR

International Nuclear Information System (INIS)

Nishimura, N.; Nakai, H.; Ross, M.A.

1999-01-01

The first Advanced Boiling Water Reactor (ABWR) started commercial operation as Tokyo Electric Power Company's (TEPCO) Kashiwazaki-Kariwa Nuclear Power Station Unit No.6 (K-6) in November 1996 and its sister Unit No.7 (K-7) in July 1997. The ABWR was developed to achieve higher reliability and safety margin while improving overall operability and economics. To achieve these goals, the optimal Boiling Water Reactor (BWR) technologies had been studied, tested and were finally adopted into the ABWR design. These technologies were called 'First of a Kind' and include the Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), and integrated digital Instrumentation and Control System (I and C). Intensive development study, confirmation tests and verification tests were conducted by the plant equipment suppliers, electric utilities, and government agencies. During plant construction, the system and equipment functions and operational characteristics were confirmed through various tests. Before the start of the commercial operation, it was confirmed that these first ABWR units met all the design goals that had been established for the next generation of nuclear power plants. Both units have now completed the first fuel cycle of operation without any unplanned plant shut down and have achieved very high availability factors. This paper describes the development and construction of these first ABWR units in the world. (author)

Shutdown risk monitoring in TEPCO

International Nuclear Information System (INIS)

Sato, Hiroki; Masuda, Takahiro; Denda, Yasutaka; Yoneyama, Mitsuru; Imai, Shun-ichi; Miyata, Koichi

2009-01-01

At present, we are introducing risk monitors into our all three nuclear power stations; Fukushima Daiichi, Fukushima Daini and Kashiwazaki Kariwa, with technical support of TEPSYS. By monitoring shutdown risk of each unit, we are trying to optimize risks during outage inspection, and raising staff's awareness for reactor safety. This paper presents our recent shutdown risk monitoring activities in Fukushima Daiichi NPS. Shutdown risk monitoring has been carried out for the past five outages of Fukushima Daiichi NPS. Daily-changing shutdown risk is evaluated in the form of core damage frequency (CDF [/day/reactor]). We also examine high-risk point of outage plan if CDF is greater than the threshold at anytime of outage. The results are delivered to operational and maintenance staff before outage. The threshold value is set ten times as much as CDF of unit in operation. As CDF exceeds the threshold, we try to either change the system configuration, or let workers pay more attention to their works during the high-risk period. We already have some examples of outage plan modification to reduce CDF using the risk monitoring information. Greater number of station staff tends to pay more attention to shutdown risk thanks to these activities. (author)

Development of alarm handling methods for boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Yukiharu, Ohga; Hiroshi, Seki; Setsuo, Arita [Power and Industrial Systems R and D Div., Hitachi, Ltd., Hitachi, Ibaraki (Japan)

1997-09-01

A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators` demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators` demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs.

Development of alarm handling methods for boiling water reactors

International Nuclear Information System (INIS)

Ohga Yukiharu; Seki Hiroshi; Arita Setsuo

1997-01-01

A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators' demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators' demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs

Radiation exposure reduction technologies for a Japanese Advanced BWR (Dose Rate Reduction Experience in Shika Unit 2)

International Nuclear Information System (INIS)

Saito, Takeshi; Ichikawa, Koji; Ishimaru, Hiroshi; Aizawa, Motohiro; Sato, Yoshiteru; Morita, Shoichi

2012-09-01

Operating experiences of the advanced boiling water reactor (ABWR) have been accumulated in Japan since the first ABWRs Kashiwazaki-Kariwa NPS unit 6 and 7 came into service. Shika NPS unit 2 (Shika-2) of Hokuriku Electric Power Co. is the fourth ABWR plant in Japan. Since ABWRs have no piping of the reactor recirculation system (RRS), which is the largest source of radiation in conventional BWRs, carbon steel piping of the reactor water cleanup system (RWCU) and residual heat removal system (RHR) are the largest source in ABWRs. Therefore we have focused on reduction methods of radioactive material on carbon steel surface in order to reduce the quantity of occupational exposure in Shika-2. In Shika-2 the following methods have been adopted. Before fuel loading, alkaline pre-filming process was applied to the RWCU piping during plant startup testing. After start of operation, the feed water iron concentration control (or nickel/iron ratio control) method was applied. Furthermore, during shutdown operation the RHR system was operated when the reactor water temperature was dropped to 120 degree C with the use of condenser heat sink cooling operation. These dose rate reduction methods worked well in Shika-2 as expected. The quantity of occupational exposure at the 3 rd outage in Shika-2 was approximately 0.35 person-Sv. On the basis of the results obtained thus far, the occupational dose expected at the outage after deposition amount of radioactivity reaching the equilibrium state has been estimated to be around 0.5 person-Sv. This value is considered to be in low level compared with the worldwide statistics. (authors)

Ground motion characteristics of 2007 Niigata-ken Chuetsu-oki earthquake

International Nuclear Information System (INIS)

Hijikata, Katsuichirou; Nishimura, Isao; Mizutani, Hiroyuki; Tokumitsu, Ryoichi; Mashimo, Mitsugu; Tanaka, Shinya

2010-01-01

Strong motion records of 2007 Niigata-ken Chuetsu-oki earthquake were examined in order to evaluate ground motion characteristics of the earthquake. Ground motions observed at Kashiwazaki Kariwa Nuclear Power Plant site were significantly larger than the response spectra evaluated on the basis of Noda et al. (2002), and the level of the ground motion observed at Arahama area (unit 1-4 side) was approximately twice as large as that at Ominato area (unit 5-7 side). Observation records of the offshore events other than the earthquake were also larger than the response spectra based on Noda et al. (2002), whereas records of the inland events were smaller than those. In addition, these characteristics were also observed in the vicinity of the site through the analysis of the ground motion records obtained by KiK-net. (author)

Human-system interfaces for the ABWR control room

International Nuclear Information System (INIS)

Gutierrez, R.; O'Neil, T.J.

1993-01-01

The 1300-MW(electric) advanced boiling water reactor (ABWR) design has been developed by General Electric Co. (G.E.) and its technical associates, Hitachi Ltd. and Toshiba Corporation, under the sponsorship of the Tokyo Electric Power Company. Key features of the ABWR include simplification, improved safety and reliability, reduced occupational exposure and radwaste, improved maneuverability, and reduced construction, fuel, and operating costs. The ABWR incorporates the best proven features from BWR designs in Japan, Europe, and the United States, combined with the broad-scope application of leading-edge technology. The first application of the ABWR design will be at TEP-CO's Kashiwazaki-Kariwa nuclear power station units 6 and 7, which are scheduled to begin commercial operation in 1996 and 1997, respectively. The ABWR is also the lead plant in the U.S. standard plant design certification program. It is scheduled to receive final approval from the U.S. Nuclear Regulatory Commission in early 1994. The development of the human system interaces (HSI) for the ABWR is an extension of the operating experience of existing man-machine interfaces currently being used at GE-designed plants

Development of next BWR plant

International Nuclear Information System (INIS)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

1995-01-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

Development of next BWR plant

Energy Technology Data Exchange (ETDEWEB)

Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

1995-04-01

It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

Social acceptance for seismic safety of nuclear installations

International Nuclear Information System (INIS)

Oiso, Shinichi

2010-01-01

The social acceptance of seismic safety of the nuclear installations was considered based on the situation that people's concern and anxieties for it having risen by earthquake suffering of the Kashiwazaki Kariwa facility in 2007, etc. It aimed mainly to extract a social awareness (acknowledgment and evaluation) which is peculiar to the earthquake in the field of nuclear power generation, and to show the attention point concerning the public relations of seismic safety of the nuclear power plant. As a result, it was suggested that we should explain based on the opinion of the third party which has a high trust such as specialist scholars, and emphasize that the severe examinations of outside third parties such as committee of the prefecture are conducted. (author)

Quality control of concrete in construction of No.2 and No.5 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc

International Nuclear Information System (INIS)

Yamashita, Toshio; Nukui, Yasushi; Nojiri, Takaharu.

1990-01-01

The construction site and its weather condition are outlined. In the Nuclear Power Station, No.1 plant and No.5 plant with 1100 MWe output each have been already in operation, and No.2 plant with 1100 MWe output is about to start the operation. In this report, the quality control of about 700,000 m 3 of concrete used for No.2 and No.5 plants construction from October, 1983 to November, 1989 is described. The features of the concrete used are shown. The facilities for producing the concrete, which were set up in the construction site, the mixing of the concrete and the quality control are reported. The system for carrying out the quality control of materials and concrete is shown with a flowchart. The material testing on cement, aggregate, water quality, fly ash and chemical additives was carried out. The slump, air quantity, temperature, strength and specific gravity of concrete were tested. (K.I.)

Emergency response of Fukushima Daini Nuclear Power Station during the Great East Japan Earthquake and its lessons

International Nuclear Information System (INIS)

Kawamura, Shinichi

2016-01-01

At the time of the occurrence of the Great East Japan Earthquake, Fukushima Daini Nuclear Power Station (hereinafter, Fukushima Daini) was operating four units of BWRS-5 type plants with an output of 1,100 MWe/unit. Among these plants, No. 1, 2, and 4 Units lost all the functions of heat removal equipment of reactors affected by tsunami. However, ad-hoc activities such as the exchange of submerged motors and temporary power installation allowed the recovery of residual heat removal (RHR) system, leading to a success in cold shutdown. This is a success story more than expectation in dealing with emergency situations, but not necessarily all of the correspondences were successfully carried out, leaving some problems. As lessons, the following are pointed out: (1) confirmation of the damage situation of the site and setting of priority rank of recovery, (2) securement of the means that do not depend on initial on-site activities, and (3) possession at the site of the skills of emergency restoration, equipment diagnostic technology, and repair technology. With reflecting lessons and challenges in these correspondences, Tokyo Electric Power Company is working to improve the accident response capability of the organization including Kashiwazaki-Kariwa Nuclear Power Station. As an example of effort of emergency response capability strengthening, there is an application of the US Incident Management System (IMS). The company is continuously making efforts for improving safety through training. (A.O.)

Concept of electric power output control system for atomic power generation plant utilizing cool energy of stored snow

International Nuclear Information System (INIS)

Kamimura, Seiji; Toita, Takayuki

2003-01-01

A concept of the SEAGUL system (Snow Enhancing Atomic-power Generation UtiLity) is proposed in this paper. Lowering the temperature of sea water for cooling of atomic-power plant will make a efficiency of power generation better and bring several ten MW additional electric power for 1356 MW class plant. The system concept stands an idea to use huge amount of seasonal storage snow for cooling water temperature control. In a case study for the Kashiwazaki-Kariwa Nuclear Power Station, it is estimated to cool down the sea water of 29degC to 20degC by 80 kt snow for 3 hours in a day would brought 60 MWh electric power per a day. Annually 38.4 Mt of stored snow will bring 1800 MWh electric power. (author)

Environmental radiation monitoring system in nuclear power station

International Nuclear Information System (INIS)

Matsuoka, Sadazumi; Tadachi, Katsuo; Endo, Mamoru; Yuya, Hiroshi

1983-01-01

At the time of the construction of nuclear power stations, prior to their start of operation, the state of environmental radiation must be grasped. After the start of the power stations, based on those data, the system of environmental radiation monitoring is established. Along with the construction of Kashiwazaki-Kariwa Nuclear Power Station, The Tokyo Electric Power Co., Inc. jointly with Fujitsu Ltd. has developed a high-reliability, environmental radiation monitoring system, and adopted ''optical data highways'' using optical fiber cables for communication. It consists of a central monitoring station and 11 telemeter observation points, for collecting both radiation and meteorological data. The data sent to the central station through the highways are then outputted on a monitoring panel. They are analyzed with a central processor, and the results are printed out. (Mori, K.)

Review of Seismic Evaluation Methodologies for Nuclear Power Plants Based on a Benchmark Exercise

International Nuclear Information System (INIS)

2013-11-01

Niigataken-chuetsu-oki (NCO) earthquake (Mw = 6.6) occurred on 16 July 2007 and affected the Kashiwazaki-Kariwa (K-K) NPP in Japan. Although there was significant loss of main shock data due to transmission problems, a significant number of instruments were still able to measure the acceleration at different locations in soil (boreholes) and in structures at the K-K NPP during the main shock and the aftershocks. The availability of all these instrumental data provided an excellent background for initiating a benchmarking exercise known as the KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment (KARISMA). The main objective of the KARISMA benchmark exercise is to study a comparison between analytical seismic response versus real response of selected structure, system and components (SSCs) of K-K NPP Unit 7. The KARISMA benchmark exercise includes benchmarking the analytical tools and numerical simulation techniques used for predicting seismic response of NPP structures (in linear and non-linear ranges), site response, soil-structure interaction phenomena, seismic response of piping systems, 'sloshing' in the spent fuel pool and buckling of tanks. The benchmark is primarily based on data provided by Tokyo Electric Power Company (TEPCO). It is not linked to the seismic re-evaluation of K-K NPP carried out by TEPCO. Twenty-one organizations, comprising researchers, operating organizations, regulatory authorities, vendors and technical support organizations from 14 countries, participated in the benchmarking exercises. This publication, including a CD-ROM, summarizes the analyses of the main results of the benchmarking exercise for the K-K NPP reactor building (including static and modal analyses of the fixed base model, soil column analyses, analyses of the soil-structure models and margin assessment of the K-K NPP reactor building), the analyses of the main results of the benchmarking exercise for the residual heat removal piping system (including

Construction of System for Seismic Observation in Deep Borehole (SODB) - Overview and Achievement Status of the Project

International Nuclear Information System (INIS)

Kobayashi, Genyu

2014-01-01

The seismic responses of each unit at the Kashiwazaki-Kariwa NPP differed greatly during the 2007 Niigata-ken Chuetsu-oki Earthquake; the deep sedimentary structure around the site greatly affected these differences. To clarify underground structure and to evaluate ground motion amplification and attenuation effects more accurately in accordance with deep sedimentary structure, JNES initiated the SODB project. Deployment of a vertical seismometer array in a 3000-meter deep borehole was completed in June 2012 on the premises of NIIT. Horizontal arrays were also placed on the ground surface. Experiences and achievements in the JNES project were introduced, including development of seismic observation technology in deep boreholes, site amplification measurements from logging data, application of borehole observation data to maintenance of nuclear power plant safety, and so on. Afterwards, the relationships of other presentations in this WS, were explained. (authors)

Preliminary Assessment for the Effects of the External Hazard Factors on the Safety of NPPs

International Nuclear Information System (INIS)

Jin, So Beom; Hyun, Seung Gyu; Kim, Sang Yun; Lee, Sung Kyu; Hur, Youl

2010-01-01

The Ch.etsu Offshore Earthquake(2007.7.16) in Japan caused damage to the Kashiwazaki-Kariwa(K-K) Nuclear Power Plants (NPPs) with seismic ground motion that exceeded the design level. This incident drew the interest of the safety evaluation studies for NPPs subjected to earthquakes exceeding the design basis around the world. Also, the Indian Ocean Tsunami(2004.12.26) tripped the Madras NPP by reason of flooding of the intake pump house and inundated the construction site of a fast breeder reactor site in India. In addition, from the various man-made and natural hazards such as the oil spill accident near Mallipo, Taean, Chungnam (2007.12), the forest fire near the Ulchin NPP site, the several inflows of marine organism into the intake of the Ulchin NPP, it was confirmed that the safety of NPPs may be affected by natural and human induced disasters. Intergovernmental Panel on Climate Change (IPCC) has been warned about global warming; the average temperature rose about 1.5 .deg. C during the 20th century and the damages caused by typhoons and heavy rains have also increased in Korea. Accordingly, a natural disaster prevention research team(hereafter team) ,which have been organized and operated since 2009, has assessed the impact of various hazards such as earthquakes and environmental changes due to global warming on the safety of NPP and has discussed to establish countermeasures. This paper introduces that the preliminary assessment for the effects of the external hazard factors on the safety of NPPs was conducted by the team

Application of ABWR construction database to nuclear power plant project

International Nuclear Information System (INIS)

Takashima, Atsushi; Katsube, Yasuhiko

1999-01-01

Tokyo Electric Power Company (TEPCO) completed the construction of Kashiwazaki-Kariwa Nuclear Power Station Unit No. 6 and No. 7 (K-6/7) as the first advanced boiling water reactors (ABWR) in the world successfully. K-6 and K-7 started their commercial operations in November, 1996 and in July, 1997 respectively. We consider ABWR as a standard BWR in the world as well as in Japan because ABWR is highly reputed. However, because the interval of our nuclear power plant construction is going to be longer, our engineering level on plant construction will be declining. Hence it is necessary for us to maintain our engineering level. In addition to this circumstance, we are planning to wide application of separated purchase orders for further cost reduction. Also there is an expectation for our contribution to ABWR plant constructions overseas. As facing these circumstances, we have developed a construction database based on our experience for ABWR construction. As the first step of developing the database for these use, we analyzed our own activities in the previous ABWR construction. Through this analysis, we could define activity units of which the project consists. As the second step, we clarified the data which are treated in each activity unit and the interface among them. By taking these steps, we could develop our database efficiently. (author)

How to press-release and take care of expression in articles at press side

International Nuclear Information System (INIS)

Hamamoto, Kazuko; Narabayashi, Tadashi; Kobayashi, Masahide; Akizuki, Teruo; Onishi, Hidetoshi

2008-01-01

As for the influences of Chuetsu-Oki Earthquake on Kashiwazaki-Kariwa Nuclear Power Station, we can conclude that 'the safety function of the nuclear power station, that is, 'Shutdown', 'Cooling down' and 'Isolating' functioned as per designed even against earthquake beyond assumption, and fundamental nuclear safety could be assured'. Nevertheless, it is said that one of the causes that harmful rumor had spread was due to mass communication media. In the press reports on some nuclear power station when affected by an earthquake and on trouble in the nuclear power station in the future, we will propose that the publication should be really useful for habitants and citizens and be promoted in the positive and expected direction, in order not to make the same mistake as this time. JSME should aim at implement the above-mentioned proposal under cooperation with other academic societies and organizations. (author)

Recent control and instrumentation systems for BWR nuclear power plant

International Nuclear Information System (INIS)

Fujii, Hiroaki; Higashikawa, Yuichi; Sato, Hideyuki

1990-01-01

For the needs of the more stable operation of nuclear power stations, the upgrading of the measurement and control system for BWRs has been promoted by positively introducing remarkably advancing electronic technology. Further, it is aimed at to construct the synthetic digitized measurement and control system for nuclear power stations to heighten the operation reliability in ABWRs. As the first step of the development in the synthetic digitization, the monitoring and control system for radioactive waste treatment was put in practical use for No.5 plant of Kashiwazaki, Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. Hitachi Ltd. has promoted the development and the application to actual plants of the measurement and control system for BWRs, in which digital control technology, optical information transmission technology and the operation-supporting technology using a computer were utilized. Hereafter, it is intended to expand the application of digital measurement and control aiming at improving the reliability, operation performance and maintainability. The nuclear power plant control complex with advanced man-machine interface-90 (NUCAMM-90) was developed, and its application to actual plants is planned. (K.I.)

Adoption of digital safety protection system in Japan

International Nuclear Information System (INIS)

Ogiso, Z.

1998-01-01

The application of micro-processor-based digital controllers has been widely propagated among various industries in recent years. While in the nuclear power plant industry, the application of them has also been expanding gradually starting from non-safety related systems, taking advantage of their reliability and maintainability over the conventional analog devices. Based on the careful study of the feasibility of digital controllers to the safety protection system, the Tokyo Electric Power Company proposed on May 1989 the adoption of digital controllers to the safety protection system in the Application for Permission of Establishment of Kashiwazaki-Kariwa units 6 and 7 (ABWR-1350Mwe each). MITI, Ministry of International Trade and Industry, the Japanese regulatory body for electric power generating facilities, had approved this application after careful review. This paper describes a series of supporting activities leading to the MITI's approval of the digital safety protection system and the MITI's licensing activities. (author)

Tsunami Research Status in IAEA after Fukushima Event

International Nuclear Information System (INIS)

Kim, Min Kyu; Choi, In Kil

2012-01-01

On March 11 th , 2011, a tremendous earthquake and tsunami occurred on the east coast of Japan. This 9.0 magnitude earthquake was the fifth greatest earthquake ever experienced on the planet. The most remarkable problem was that the Fukishima NPP sites. After Japan earthquake, many international researches about tsunami and earthquake event were started or revised. Especially, the most remarkable point of the great earthquake in east coast of Japan was tsunami event. Before this earthquake, the Niigata earthquake occurred in 2007 and the Kashiwazaki Kariwa nuclear power plant had little damaged. The research about the safety of nuclear power plant against earthquake events was activated by 2007 Niigata earthquake. However, the researches about a tsunami event were very few and only tsunami simulation was only focused. After the Fukushima accident, the international society became very interested in tsunami event as a major external event. Therefore in this study, the tsunami research status in IAEA after Fukushima event and the role of Korea are introduced

Outline of Fukushima nuclear accident and future action. Lessons learned from accident and countermeasure plan

International Nuclear Information System (INIS)

Fukuda, Toshihiko

2012-01-01

Fukushima nuclear accident was caused by loss of all AC power sources (SBO) and loss of ultimate heat sink (LUHS) at Fukushima Daiichi Nuclear Power Plants (NPPs) hit by the Great East Japan Earthquake. This article reviewed outline of Fukushima nuclear accident progression when on year had passed since and referred to lessons learned from accident and countermeasure plan to prevent severe accident in SBO and LUHS events by earthquake and tsunami as future action. This countermeasure would be taken to (1) prevent serious flooding in case a tsunami overwhelms the breakwater, with improving water tightness of rooms for emergency diesel generator, batteries and power centers, (2) enhance emergency power supply and cooling function with mobile electricity generator, high pressure fire pump car and alternate water supply source, (3) mitigate environmental effects caused by core damage with installing containment filtered venting, and (4) enforce emergency preparedness in case of severe accident. Definite countermeasure plan for Kashiwazaki-Kariwa NPPs was enumerated. (T. Tanaka)

Development of Innovative Construction Technologies for ABWRs

International Nuclear Information System (INIS)

Yamashita, T.; Fukuda, T.; Masuda, N.

2002-01-01

This paper describes an effort by Tokyo Electric Power Company (TEPCO) to shorten the construction time in a drastic manner for the Advanced Boiling Water Reactors (ABWR), thereby aiming at reducing construction costs. First an outline of the actual construction records for the five BWR Units and the two ABWR Units at the Kashiwazaki-Kariwa site is introduced along with the construction methods employed for these units. There is a continued trend in the reduction of construction time from Unit 1 to 7 owing to a number of improvements made to these units, and in particular it is noteworthy that the drastic reduction was accomplished due to the change in reactor type from BWR to ABWR. Explained next is an on-going effort for the next ABWR and the following next generation of ABWRs to further shorten construction time. In this effort an emphasis is laid on the development of innovative construction methods by the adoption of steel plate/concrete composite structures (SC Structure), and the application of those structures even to containment vessel (so-called SCCV). This work is being conducted as part of TEPCO's research and development for the next generation ABWRs (ABWR II). Presently it is expected that the construction time from bedrock inspection to fuel loading could be shortened in a stepwise fashion to 38 and 31.5 months for the next ABWR and finally to 23 months for the next generation ABWR II, thus enabling a greatly reduced power generating cost and enhanced safety during construction. In addition to the above effort, a preliminary study has been performed concerning the application of base isolation systems to ABWR plants. From this study which is aimed at standardization, lowered cost and mitigation of engineering work, it was found that base isolation system effectively reduces seismic response and is instrumental in achieving the above-mentioned objectives. Therefore it can be said that the use of this system gives more freedom in selecting sites

Report of fact-finding survey in atomic energy industries, 1990

International Nuclear Information System (INIS)

Anon.

1992-01-01

In fiscal year 1990, Japanese economy maintained steady domestic civilian demands, and the real rate of economical growth became 5.7%. The total supply of primary energy increased by 5.3% as compared with that in the last year. The share of nuclear power recovered to 9.4% due to the increase of the capacity factor after the lapse of three years. The demand of electric power increased by 7.3% as compared with that in the previous year, and the share of nuclear power in the total generated electric power was 26.3%. The outlay of electric power industry related to atomic energy was 1735.4 billion yen, and 6% increase as compared with that in the previous year. The sales of mining and manufacturing industries related to atomic energy was 1768.1 billion yen and 2% increase as compared with that in the previous year. The construction cost of nuclear power stations was 658.3 billion yen and 17% increase. The civilian workers related to atomic energy in electric power, mining and manufacturing industries were 57687 persons and 7% increase. Kashiwazaki-Kariwa No. 2 plant (1100 MWe BWR) and No. 5 plant (1100 MWe BWR) of Tokyo Electric Power Co., Inc. started the operation, and installed nuclear power generation capacity reached 31.48 million kW of 39 plants. (K.I.)

Study on post-earthquake plant evaluation and communication (Annual safety research report, JFY 2011)

International Nuclear Information System (INIS)

2012-01-01

The aims of this study are to establish a post-earthquake plant evaluation method and to develop a communication system for the improving seismic safety regulations as well as encouraging public communication. The Miyagiken-oki earthquake in 2005, Onagawa Nuclear Power Plant shut down automatically. Subsequently, JNES started development of post-earthquake plant evaluation and communication system based on the experience of the cross-check analysis for Onagawa Nuclear Power Plant. The Niigata-ken Chuetsu-oki Earthquake in 2007, the plant situation was not transmitted promptly. The loss of information sharing between local community and related organizations caused the public anxiety. The importance of plant information transmission as well as seismic information gathering were recognized. The proposal for the solution of the information issues were performed by government committee, In this study the evaluation method for soundness of the main structure and equipment after earthquake event were updated. Moreover, information dissemination to the public and transparent communication methodology was examined by the Industry-Academia-Government cooperation in the Kashiwazaki-Kariwa region. (author)

Innovation in civil construction system of nuclear power plant

International Nuclear Information System (INIS)

Takami, Masahiro

1996-01-01

Nowadays, the computer-aided production systems have been already introduced to almost all kinds of industries. The construction industry, which has been said to be conservative for the modernization of production system, now expects the CIC (Computer Integrated Construction) as the means to innovate the construction production process. Shimizu Corporation has developed the new computer-aided production system, 'SIPS: Shimizu Integrated Production System', and has used it in the actual construction projects. In the system, the computer supports every phase of construction projects like market researching, design, material purchase, construction work, and maintenance. The project of Kashiwazaki-kariwa Nuclear Power Station Unit No.7 is one of the model cases. Here we applied following three concepts, (1) the full use and integration of 3D-CAD data-base through all phases of construction, (2) the setting-up of the information network system among the site office, the head office, and the mechanical and electrical manufacturer, (3) the introduction of advanced construction technologies such as large block prefabrication method. (author)

Operating Experience with the BR-5 Reactor; Experience acquise aupres du reacteur BR-5; Opyt ehkspluatatsii reaktora BR-5; Experiencia practica con el reactor BR-5

Energy Technology Data Exchange (ETDEWEB)

Lejpunskij, A. I.; Kazachkovskij, O. D.; Pinkhasik, M. S.; Aristarkhov, N. N.; Karpov, A. V.; Larin, E. P.; Efimov, I. A.

1963-10-15

The paper discusses the carrying-out of repair and maintenance work on the radioactive liquid-metal circuit of the BR-5 fast neutron reactor. Attention is also given to problems of reactor operation after achievement of the planned 2% fuel burn-up with some disturbance of leak-tightness in individual fuel elements. An account is given of experience in discharging the active section, examining the condition and leak-tightness of the fuel elements, and decontaminating the equipment and piping of the first radioactive circuit after reaching 5% fuel burn-up. (author) [French] Dans ce memoire les auteurs decrivent l'execution des reparations et des travaux d'entretien dans le circuit radioactif liquide-metal du reacteur a neutrons rapides BR-5. Ils etudient egalement les problemes lies au fonctionnement du reacteur au taux de combustion de 2% prevu avec quelques defauts d'etancheite dans des elements combustibles particuliers. Ils decrivent le dechargementen zone active et examinent les conditions d'etancheite des elements combustibles. Ainsi que la decontamination de l'appareillage et des tuyauteries du premier circuit radioactif apres avoir atteint un taux de combustion de 5%. (author) [Spanish] En la memoria se examinan los problemas planteados por el mantenimiento del circuito radiactivo de metal liquido del reactor de neutrones rapidos BR-5. Se tratan cuestiones relacionadas con la explotacion del reactor una vez alcanzado el grado de combustion de 2%, previsto en el proyecto y luego de producirse ciertas alteraciones de la densidad de determinados elementos combustibles. Se describen la experiencia adquirida durante la descarga del cuerpo del reactor, las investigaciones del estado general y de la hermeticidad de los elementos combustibles y las operaciones de descontaminacion de la instalacion y de las tuberias del circuito radiactivo primario despues de alcanzado un grado de combustion de 5%. (author) [Russian] V doklade rassmatrivayutsya voprosy proizvodstva

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Overview of Nuclear Fuel-Cycle Policy and the Role of the Nuclear Safety Commission in Japan

International Nuclear Information System (INIS)

Higashi, K.; Nishinosono, S.

2008-01-01

Since the first generation of electricity by the Japan Power Demonstration Reactor in 1963, Japan has been extensively developing nuclear technologies solely for peaceful purposes. The country now operates 55 nuclear power plants consisted of BWRs and PWRs. Although Japan is one of the largest consumers of energy in the world, the country has very limited domestic energy resources. Therefore, Japan considers the nuclear power generation very important as plutonium and uranium recovered from spent fuels can be used in new nuclear fuels as quasi-domestic energy resource. For recycle use of nuclear fuels, the establishment of nuclear fuel recycling technologies including reprocessing technologies is essential. Since 1977, Japan has been recovering plutonium and uranium by a small scale reprocessing plant built by French technology. Recently, 800 ton/year scale commercial reprocessing plant is under construction. After overcoming the current technical problem in the vitrification facility, the commercial plant is expected to be in full operation soon. Concerning the disposal of radioactive wastes, which arises from nuclear utilization, sallow land disposal has already been implemented and medium depth (50 to 100 m) disposal plan is in progress. For high-level waste, possible candidate sites for disposal are being sought. In this paper, the statuses of nuclear power plants and of nuclear fuel cycle facilities in Japan are summarized. As safety is essential for these nuclear installations, safety regulations in Japan are briefly presented from the viewpoint of Nuclear Safety Commission. Furthermore, as the most significant recent safety issue in Japan, the impacts of the large near-site earthquake hit Kashiwazaki-Kariwa NPP last July are reported.(author)

Study of applicable methods on safety verification of disposal facilities and waste packages

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Three subjects about safety verification on the disposal of low level radioactive waste were investigated in FY. 2012. For radioactive waste disposal facilities, specs and construction techniques of covering with soil to prevent possible destruction caused by natural events (e.g. earthquake) were studied to consider verification methods for those specs. For waste packages subject to near surface pit disposal, settings of scaling factor and average radioactivity concentration (hereafter referred to as ''SF'') on container-filled and solidified waste packages generated from Kashiwazaki Kariwa Nuclear Power Station Unit 1-5, setting of cesium residual ratio of molten solidified waste generated from Tokai and Tokai No.2 Power Stations, etc. were studied. Those results were finalized in consideration of the opinion from advisory panel, and publicly opened as JNES-EV reports. In FY 2012, five JNES reports were published and these have been used as standards of safety verification on waste packages. The verification method of radioactive wastes subject to near-surface trench disposal and intermediate depth disposal were also studied. For radioactive wastes which will be returned from overseas, determination methods of radioactive concentration, heat rate and hydrogen generation rate of CSD-C were established. Determination methods of radioactive concentration and heat rate of CSD-B were also established. These results will be referred to verification manuals. (author)

World class BWRs for the 1990s

International Nuclear Information System (INIS)

Wilkins, D.; Lowton, R.

1989-01-01

The first step in Advanced BWR development took place in 1978. It was the formation in San Jose, California, of an Advanced Engineering Team, consisting of representatives of worldwide BWR suppliers and leading architect engineer firms. The Advanced Engineering Team effort was completed in 1979 and established the basic ABWR concept. During the period 1980-85, ABWR development was focused in Japan. Under the sponsorship of the Tokyo Electric Power Company and Japanese BWR utilities, GE, Hitachi and Toshiba performed design development studies to confirm the technical and economic attractiveness of the ABWR, and also conducted a comprehensive test programme encompassing the new features. In 1987, The Tokyo Electric Power Company announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa nuclear power station with commercial operation of the first unit in 1996 and the second in 1998. The company submitted the application for the establishment permit for the two ABWR units in May 1988. The units will be supplied by a joint venture of the three companies who worked on the development studies. This article goes on to describe the ABWR design resulting form this international development effort. (3 figures, 1 table). (Author)

Non-linear behavior of public opinion on the issues regarding inhabitants' polls

International Nuclear Information System (INIS)

Ohnishi, Teruaki; Suganuma, Jyun-ichi

2003-01-01

The observed change of public attitude around the time of inhabitants' poll in Japan was compared with model calculation to investigate its non-linear behavior. Two inhabitants' polls regarding nuclear issues, the approval and disapproval of the construction of Maki nuclear station, and of the MOX fuel use at Kashiwazaki-Kariwa nuclear station, were considered together with the poll on the reconstruction of the tenth weir in Yoshino river carried out in Tokushima. By using a mathematical model such that the individual attitude is mainly subject to two factors of the information environment and the mutual communication between the public, it was found that the change and the unification of public attitude around the time of inhabitants' poll can be interpreted as a manifestation of self-organization resulted from the cooperative phenomenon of those two factors. Moreover, it was also found that the abrupt change of public attitude just before the poll can be interpreted as a result of positive feedback of the information environment formed by the various types of propaganda activities to the attitude change, though the extent of such non-linear effects differs from case to case. (author)

NS [Nuclear Safety] update. Current safety and security activities and developments taking place in the Department of Nuclear Safety and Security, Issue no. 7, July 2008

International Nuclear Information System (INIS)

2008-09-01

The current issue presents information about the following activities: 1) International Conference on Radioecology and Environmental Radioactivity 15-20 June 2008, Bergen organized by the Norwegian Radiation Protection Authority (NRPA) and the French Institute for Radiation Protection and Nuclear Safety (IRSN) in cooperation with IAEA, WHO, OECD/NEA, the International Union of Radioecology (IUR), the International Commission on Radiological Protection (ICRP) and the Journal of environmental Radioactivity (JER). This conference has evolved from the merging of the series of conferences by ECORAD and of the International Conferences on Radioactivity in the Environment. The primary objective of the conference was to review all scientific themes related to the study of environmental radioactivity and to identify new societal needs and requirements for regulatory bodies and industry. All sources of radiation, from industrial discharges in planned exposure situations through to sources in existing and emergency exposure situations, were included in the scope of the conference. 2) The Ibero American Forum of Nuclear and Radiation Safety and Security Regulatory Agencies (the FORO); 3) Kashiwazaki-Kariwa Nuclear Power Plant in the earthquake on 16 July 2007 - plant's response and effects and lessons learned

Effectiveness for public acceptance of nuclear energy. Analysis of questionnaire distributed at organized visits to Nuclear Power Stations

International Nuclear Information System (INIS)

1996-01-01

In First Atomic Power Industry Group, as a part of the public acceptance activities for nuclear power, the visit to nuclear power stations for the inspection have been held. This aims at acquiring the correct knowledge on nuclear power by seeing the various facilities of nuclear power stations actually by own eyes. In the execution of the visits, mainly female employees were asked to participate, calling to the various companies in the group. On November 16 and 17, 1995, the visit to Kashiwazaki-Kariwa nuclear power station of Tokyo Electric Power Co. was held. In this study, the impression and opinion that all the participants possess through the visit and the seminar held beforehand were collected and analyzed by the questionnaire. The schedule of the visit, and the contents of the questionnaire are shown. As the feature of the questionnaire, the same questions were made before and after the inspection, and the change of opinion was examined. The results are summarized. The participants who trust the safety of nuclear power stations were 37% before the inspection and increased to 84% after the inspection. (K.I.)

Digital electro-hydraulic control system for nuclear turbine

International Nuclear Information System (INIS)

Yokota, Yutaka; Tone, Youichi; Ozono, Jiro

1985-01-01

The unit capacity of steam turbines for nuclear power generation is very large, accordingly their unexpected stop disturbs power system, and the lowering of their capacity ratio exerts large influence on power generation cost. Therefore, very high reliability is required for turbine EHC controllers which directly control the turbines for nuclear power generation. In order to meet such requirement, Toshiba Corp. has developed high reliability type analog tripled turbine EHC controllers, and delivered them to No. 3 plant in the Fukushima No. 2 Nuclear Power Station and No. 1 plant in the Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. At present, the trial operation is under way. The development of digital EHC controllers was begun in 1976, and through the digital EHC for a test turbine and that for a small turbine, the digital EHC controllers for the turbines for nuclear power generation were developed. In this paper, the function, constitution, features and maintenance of the digital tripled EHC controllers for the turbines for nuclear power generation, the application of new technology to them, and the confirmation of the control function by simulation are reported. (Kako, I.)

FAPIG's activities for public acceptance of nuclear energy. Analytical results of questionnaire executed at organized visits to nuclear power stations

International Nuclear Information System (INIS)

Mizoguchi, Tadao

1999-01-01

FAPIG organizes a visit to nuclear power station in every November. It is an object that visitors acquire the correct knowledge of nuclear power by looking at the various facilities in the nuclear power stations. The paper showed the analytical results of questionnaire executed at organized visits to the Kashiwazaki-Kariwa nuclear power station. The visitors were 18 women. The questionnaire was carried out by the same problems before and after seminar and a conducted tour. Their impressions and opinions and the changes are analyzed. The speakers used easy words, video, OHP, pamphlet and experimental equipment. These means showed very good results to visitors. The seminar had very large effect on just recognition of safety and need of it. The change of answer proved from 3 to 6 of need and from 0 to 7 of safety of it. Nine members indicated good understanding of seminar content. The interested items in the seminar were measurement of radiation, effects of radiation, reason of decreasing average life, Chernobyl accident, difference between nuclear power and atomic bomb and nuclear power dose not generate carbon dioxide and recycle plutonium after nuclear fission of uranium. (S.Y.)

Total system for manufacture of nuclear vessels by computer: VECTRON

International Nuclear Information System (INIS)

Inagawa, Jin; Ueno, Osamu; Hanai, Yoshiharu; Ohkawa, Isao; Washizu, Hideyuki

1980-01-01

VECTRON (Vessel Engineering by Computer Tool and Rapid Operating for the N/C System) is a CAM (Computer Aided Manufacturing) system that has been developed to produce high quality and highly accurate vessels for nuclear power plants and other industrial plants. Outputs of this system are design drawings, manufacturing information and magnetic tapes of the N/C marking machine for vessel shell plates including their attachments. And it can also output information at each stage of designing, marking, cutting, forming and assembling by treating the vessels in three dimensions and by using data filing systems and plotting program for general use. The data filing systems consist of functional and manufacturing data of each part of vessels. This system not only realizes a change from manual work to computer work, but also leads us to improve production engineering and production jigs for safety and high quality. At present, VECTRON is being applied to the manufacture of the shell plates of primary containment vessels in the Kashiwazaki-Kariwa Nuclear Power Station Unit 1 (K-1) and the Fukushima Daini Nuclear Power Station Unit 3 (2F-3), to realize increased productivity. (author)

Further improvement of human-machine interface for ABWR main control room

International Nuclear Information System (INIS)

Makino, S.

2001-01-01

Tokyo Electric Power Company (TEPCO) has developed main control room panels based on progress in C and I technology. ABWR type main control room panels (ABWR MCR PNLs) are categorized as third generation type domestic BWR MCR, that is, they are were developed step by step based on operating experience with the first and the second generation BWR. ABWR type main control room panels were applied to Kashiwazaki-Kariwa Nuclear Power Station Units Number 6 and 7 (K-6/7) for the first time. K-6/7 are the first advanced BWR (ABWR), which started commercial operation in November 1996 and July 1997, respectively. The concept of ABWR MCR design was verified through wooden mock-up panels, start-up tests and commercial operation. Though the K-6/7 design has borne fruit, we are planning to refine and standardize the design based on the following concepts: to maintain the plant operation and monitoring style of ABWR MCR PNLs; to introduce brand-new HMI technology and devices; to incorporate operators' advice in the design. This paper outlines the features and improvements of the K6/7 MCR PNLs design. (author)

Study on post-earthquake plant evaluation and communication (Annual safety research report, JFY 2010)

International Nuclear Information System (INIS)

Iijima, Toru; Taoka, Hideto; Yamada, Hiroyuki; Sano, Kyoko

2011-01-01

The aims of this study are to establish a post-earthquake plant evaluation method and to develop a communication system for the improving seismic safety regulations as well as encouraging public communication. The Miyagiken-oki earthquake in 2005, Onagawa Nuclear Power Plant shut down automatically. Subsequently, JNES started development of post-earthquake plant evaluation and communication system based on the experience of the cross-check analysis for Onagawa Nuclear Power Plant. The Niigata-ken Chuetsu-oki Earthquake in 2007, the plant situation was not transmitted promptly. The loss of information sharing between local community and related organizations caused the public anxiety. The importance of plant information transmission as well as seismic information gathering were recognized. The proposal for the solution of the information issues were performed by government committee. In this study, the evaluation method for soundness of the main structure and equipment after earthquake event were updated. Moreover, procedure of the post-earthquake plant soundness evaluation and multi-functional seismic information system were developed. In addition, the implementation strategy of the easy-to -understand information dissemination to the public and transparent communication methodology was examined by the Industry-Academia-Government cooperation in the Kashiwazaki-Kariwa region. (author)

Settlement mechanism of the backfilled ground around nuclear power plant buildings. Part 1. A series of 1G shaking table tests

International Nuclear Information System (INIS)

Ishimaru, Makoto; Kawai, Tadashi

2008-01-01

The large ground settlement locally occurred at the backfilled ground around the Kashiwazaki-Kariwa Nuclear Power Plant buildings during the Niigataken Chuetsu-oki Earthquake in 2007. The purposes of this study are to verify the assumed mechanism of the settlement and to discuss the influence factors on the settlement. For these purposes, we conducted a series of 1G shaking table tests using a rigid structure and sand. In the tests, parameters, which were variously changed, are related to two factors; one is the horizontal ground displacement relative to the structure, the other is the ground strength against the sliding failure. The following results were obtained: (1) All the results showed that the ground settlement sizes near the structure were larger than the ground settlement sizes far from the structure, (2) From the video observed at the ground near the structure, it was found that the settlement locally occurred due to the sliding failure after the ground was separated from the structure, (3) The ground settlement sizes near the structure were large as the horizontal ground displacement sizes were large, and the soil strength arising from fines affected the ground settlement sizes near the structure. (author)

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

International Nuclear Information System (INIS)

Li Tianshu

1991-09-01

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

Applicability of RELAP5 for safety analysis of AP600 and PIUS reactors

International Nuclear Information System (INIS)

Motloch, C.G.; Modro, S.M.

1990-01-01

An assessment of the applicability of using RELAP5 for performing safety analyses of the AP600 and PIUS advanced reactor concepts is being performed. This ongoing work is part of a larger safety assessment of advanced reactors sponsored by the United States Nuclear Regulatory Commission. RELAP5 models and correlations are being reviewed from the perspective of the new AP600 and PIUS phenomena and features that could be important to reactor safety. The purpose is to identify those areas in which new mathematical models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 and PIUS designs and systems and the planned and off-normal operations are similar enough to current Pressurized Water Reactors (PWR) that RELAP5 safety analysis applicability is unchanged. However, for AP600 the single most important systemic and phenomenological difference between it and current PWRs is in the close coupling between the reactor system and the containment during postulated Loss of Coolant Accident (LOCA) events. This close coupling may require the addition of some thermal-hydraulic models to RELAP5. And for PIUS, the most important new feature is the thermal density locks. These and other important safety-related features are discussed. This document presents general descriptions of RELAP5, AP600, and PIUS, describes the new features and phenomena of the reactors, and discusses the code/reactors safety-related issues. 32 refs., 4 figs., 2 tabs

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

The heating operational summarization in three winters of a 5 MW test heating reactor

International Nuclear Information System (INIS)

Wang Dazhong; Dong Duo; Su Qingshan; Zhang Yajun

1992-09-01

The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

Seismic safety reexaminations to NPPs in Taiwan. Lessons learned from 20061226 Taiwan Hengchun and 20070716 Japan Niigata-Chuetsu oki earthquakes

International Nuclear Information System (INIS)

Chow Ting; Wu Yuanchieh; Gau Yunchau

2008-01-01

On December 26 2006, a strong earthquake with a local magnitude M L of 7.0 hit the most southern part of Taiwan, Hengchun village, where the Maanshan Nuclear Power Station is located. This is a historic high earthquake ever been experienced to Taiwan's existing nuclear power units, and it raised high public concerns about the seismic safety of the nuclear power plants operation. More recently on July 16 2007, in Japan, where the earthquake focal mechanisms are very similar to those in Taiwan, all 7 nuclear power units in Kashiwazaki-Kariwa site were struck by a more devastating earthquake and as the result, the design earthquakes for all the nuclear units have been exceeded. Therefore, the assurance of good seismic design and the appropriateness of associated post-earthquake actions to the nuclear power units in Taiwan become very urgent topics. Based on the experiences learned from the above mentioned two earthquakes, this paper will focus on the seismic safety reexamination of Taiwan's existing nuclear power plants of the following aspects: (1) current US orientated seismic designs/regulations from earthquake probabilistic risk point of view, (2) earthquake shut-down criterion, especially the CAV parameter and its threshold value, and (3) current post earthquake actions. (author)

Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

International Nuclear Information System (INIS)

Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

2015-01-01

Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

Optimal operating parameters of the reactor oscillator in the channel of 6.5/10 MW reactor; Odredjivanje optimalnih radnih tacaka za reaktorski oscilator u kanalu na reaktoru 6,5/10 MW

Energy Technology Data Exchange (ETDEWEB)

Lolic, B; Zecevic, V [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1961-07-01

Operating conditions for the reactor oscillator in the central vertical experimental channel (VK5) in the RA reactor were studied during 1960. Channel VK5 was chosen because the sensitivity of the reactor is highest in this case. The central vertical experimental channel is placed in the center of the core and its bottom is placed 200 mm from the bottom of the reactor core. Diameter of the channel is is 110 mm and its length 5706 mm. During operation of the reactor oscillator with total modulation of the reactor power, it is very important to determine the oscillator operating point and the oscillation amplitude in such a way to avoid any change in reactor power level. Positive reactivity changes originating from oscillations of the samples should be compensated by the negative reactivity changes so that the effect should be nil. Operating points of the reactor oscillator are in the middle of the straight part of the figure showing the reactivity change dependent on the position of the absorber.

Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

International Nuclear Information System (INIS)

Griggs, D.P.

1993-03-01

The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

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

Robotic dismantlement systems at the CP-5 reactor D and D project

International Nuclear Information System (INIS)

Seifert, L. S.

1998-01-01

The Chicago Pile 5 (CP-5) Research Reactor Facility is currently undergoing decontamination and decommissioning (D and D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principle nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water cooled and moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D and D project includes the disassembly, segmentation and removal of all the radioactive components, equipment and structures associated with the CP-5 facility. The Department of Energy's Robotics Technology Development Program and the Federal Energy Technology Center, Morgantown Office provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor assembly for tasks requiring remote dismantlement as part of the EM-50 Large-Scale Demonstration Program (LSDP). The teleoperated systems provided were the Dual Arm Work Platform (DAWP), the Rosie Mobile Teleoperated Robot Work System (ROSIE), and a remotely-operated crane control system with installed swing-reduction control system. Another remotely operated apparatus, a Brokk BM250, was loaned to ANL by the Princeton Plasma Physics Laboratory (PPPL). This machine is not teleoperated and was not part of the LSDP, but deserves some mention in this discussion. The DAWP is a robotic dismantlement system that includes a pair of Schilling Robotic Systems Titan III hydraulic manipulator arms mounted to a specially designed support platform: a hydraulic power unit (HPU) and a remote operator console. The DAWP is designed to be crane-suspended for remote positioning. ROSIE, developed by RedZone Robotics, Inc. is a mobile, electro-hydraulic, omnidirectional platform with a heavy-duty telescoping boom mounted to the platform's deck. The work system includes the mobile platform (locomotor), a power distribution unit (PDU) and a remote operator console. ROSIE moves about the reactor building

Additional examination on station blackout caused by tsunami in Fukushima Daiichi NPS

International Nuclear Information System (INIS)

Yamauchi, Daisuke; Date, Kenji; Mizokami, Masato; Honda, Takeshi; Nozaki, Kenichiro; Mizokami, Shinya; Endo, Ryohei

2017-01-01

This study is additional examination to verify in a more reliable way the assessment that the emergency AC power supply was lost due to tsunami at Fukushima Daiichi Nuclear Power Station. It confirmed the relationship between the path length of tsunami intrusion reaching each power supply facility and the function loss time. As a result of examination, it was confirmed that as the path length of the tsunami intrusion reaching each power supply facility was longer, the function loss time tended to be later. So, conventional assessment that the function of each power supply facility was lost due to the run-up and flooding of tsunami has become more probable. For facilities, where the overall trend and the loss time of function were divergent, it was found that there were scenarios that could reasonably be explained. Based on the fact that the Fukushima Daiichi Nuclear Power Station lost power due to the tsunami, Kashiwazaki-Kariwa Nuclear Power Station carries out various safety measures. First of all, as the measures to prevent accidents caused by tsunami, the following have been applied: (1) prevention of the inflow of tsunamis into premises, (2) water prevention of the areas installed with important equipment, (3) securement of seawater at the time of backwashing, (4) storage of portable equipment at high ground, (5) installation of tsunami surveillance cameras. To prepare for the loss of power supply, this station implemented power supply facilities such as generator cars and distribution boards, as well as the placement of power supply cars at high ground. (A.O.)

Three-dimensional numerical investigation of a Molten Salt reactor concept with the code CFX-5.5

International Nuclear Information System (INIS)

Yamaji, B.; Csom, G.; Aszodi, A.

2002-01-01

Partitioning and transmutation of actinides and long-lived fission products is a promising option to extend the possibilities and enhance the environmentally acceptable capabilities of nuclear energy. Also the possible implementation of the thorium cycle is considered as a way to reduce the problem of energy resources in the future. For both objectives different molten salt reactor concepts were proposed mainly based on the Molten Salt Reactor Experiment of the Oak Ridge National Laboratory. Not only critical reactors but also accelerator-driven subcritical systems (ADSs) have advantages worth considering for those aims, especially those ones with liquid fuel, such as molten salts. By using liquid fuel which is the coolant medium, too, a basically different thermalhydraulic behavior is expected than in the case of solid fuel and water coolant. In this work our purpose is to present the possible use of Computational Fluid Dynamics (CFD) technology in molten salt thermal hydraulics. The simulations were performed with the three-dimensional code CFX-5.5.(author)

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

Scale-4 analysis of pressurized water reactor critical configurations: Volume 5, North Anna Unit 1 Cycle 5

International Nuclear Information System (INIS)

Bowman, S.M.; Suto, T.

1996-10-01

ANSI/ANS 8.1 requires that calculational methods for away-from- reactor (AFR) criticality safety analyses be validated against experiment. This report summarizes part of the ongoing effort to benchmark AFR criticality analysis methods using selected critical configurations from commercial PWRs. Codes and data in the SCALE-4 code system were used. This volume documents the SCALE system analysis of one reactor critical configuration for North Anna Unit 1 Cycle 5. The KENO V.a criticality calculations for the North Anna 1 Cycle 5 beginning-of-cycle model yielded a value for k eff of 1. 0040±0.0005

Project and characteristics of a 5MW experimental fast reactor

International Nuclear Information System (INIS)

Ishiguro, Y.; Nascimento, J.A. do.

1986-05-01

Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 10 15 n/cm 2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author) [pt

Specific issues for seismic performance of power plant equipment

Energy Technology Data Exchange (ETDEWEB)

Nawrotzki, Peter [GERB Vibration Control Systems, Berlin (Germany)

2010-01-15

Power plant machinery can be dynamically decoupled from the substructure by the effective use of helical steel springs and viscous dampers. Turbine foundations, coal mills, boiler feed pumps and other machine foundations benefit from this type of elastic support systems to mitigate the transmission of operational vibration. The application of these devices may also be used to protect against earthquakes and other catastrophic events, i.e. airplane crash, of particular importance in nuclear facilities. This article illustrates basic principles of elastic support systems and applications on power plant equipment and buildings in medium and high seismic areas. Spring damper combinations with special stiffness properties are used to reduce seismic acceleration levels of turbine components and other safety or non-safety related structures. For turbine buildings, the integration of the turbine sub-structure into the machine building can further reduce stress levels in all structural members. The application of this seismic protection strategy for a spent fuel storage tank in a high seismic area is also discussed. Safety in nuclear facilities is of particular importance and recent seismic events and the resulting damage in these facilities again brings up the discussion. One of the latest events is the 2007 Chuetsu earthquake in Japan. The resulting damage in the Kashiwazaki Kariwa Nuclear Power Plant can be found in several reports, e.g. in Yamashita. (orig.)

Effluents and releases of tritium from Novo-Voronezh-5 reactor

International Nuclear Information System (INIS)

Babenko, A.G.; Mekhedov, B.N.; Podporinova, L.E.; Popov, S.V.; Shalin, A.N.

1990-01-01

Results of systematic measurements of tritium concentration within technological systems of reactor of Novo-Voronezh NPP conducted to evaluate tritium effluents and releases and radiation doses to population from these effluents and releases are given. It is shown that 68% concerning tritium total amount were disposed into sewerage while 17% - through vent tube and 15% - with water and steam from secondary circuit systems. Standartized tritium effluents from WWER-1000 reactor for 5 year run constitute 15±1.9 GBq/MWxyear and it corresponds to mean value of effluents for foreign NPPs. Tritium concentration in the atmosphere constituted according to calculations (4.1-20)x10 -5 Bq/l. Conclusion is made about insignificant dose to population from tritium gaseous effluents. Detail study is necessary for dose connected with tritium contained in water effluents

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)

Relap5 simulation for severe accident analysis of RSG-GAS Reactor

International Nuclear Information System (INIS)

Andi Sofrany Ekariansyah; Endiah P-Hastuti; Sudarmono

2018-01-01

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational characteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element. (author)

Loss of coolant acident analyses on Osiris research reactor using the RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto Vitor; Costa, Antonella Lombardi; Lima, Claubia Pereira Bezerra; Veloso, Maria Auxiliadora Fortini

2011-01-01

RELAP5/MOD 3.3 code is widely used for thermal hydraulic studies of commercial nuclear power plants. However, several current investigations have shown that RELAP5 code can also be applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this paper, a nodalization of the core and the most important components of the primary cooling system of the OSIRIS reactor developed for RELAP5 thermal hydraulic code are presented as well as results of steady state and transient simulations. OSIRIS has thermal power of 70 MW and it is an open pool type research reactor moderated and cooled by water. The OSIRIS reactor characteristics have been used as a base for the development of a model for the Multipurpose Brazilian Reactor (RMB). The aim of the present work is to investigate the behavior of the core during a loss of coolant accident and the possible damage of the fuel elements due an inadequate heat removal. Although the core coolant reached the saturation point due the large break, the fuel element conditions were out of the damage zone. (author)

The AMPS 1.5 MW low-pressure compact reactor

International Nuclear Information System (INIS)

Hewitt, J.S.

1987-01-01

The 1.5-MWt reactor of the Autonomous Marine Power Source (AMPS) is designed to meet the unusual requirements of its first application. To provide for 100 kWe (net) on board self-sustaining manned submersible vehicles, the AMPS reactor must deliver safely, reliably and without direct operator surveillance, its thermal output to freon Rankine-cycle engines at thermodynamically useful temperatures. It must also conform to space and weight limits on the order of less than 50 cubic metres and 70 tonnes. The safety requirements are met by (i) limiting lifetime excess reactivity requirements by incorporation of burnable poison in the U-Zr-H fuel, (ii) maintaining nominal pressures in the light-water primary system at about 1 atmosphere, and (iii) maintaining a large volume of primary reserve coolant at temperature depressed relative to that of the circulating coolant. The latter averages 90 degrees celsius as it is pumped around loops that include the reactor core and the freon evaporators during normal operation. In the event of loss of pumped flow, the system defaults by intrinsic means to core cooling through natural convective exchange with the reserve coolant. In the post-shutdown situation, this passive cooling mode continues to operate regardless of vessel orientation and decay heat is safely dissipated to the sea. The design of the AMPS system, including the reactor, the freon engines, the control and monitoring system, the safety shut-down system and the power source container, are in advanced stages of design. (author)

The 5th surveillance testing for Kori unit 2 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-01

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 5th 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 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

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

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

Science.gov (United States)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

Directory of Open Access Journals (Sweden)

Khedr Ahmed

2005-01-01

Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S; Lewis, B.J.; Bonin, H.W.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

The 5th surveillance testing for Kori unit 1 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)

2000-08-01

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 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

2016-08-15

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

International Nuclear Information System (INIS)

Shi, Chengbin; Cheng, Maosong; Liu, Guimin

2016-01-01

Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

HECTR [Hydrogen Event Containment Transient Response] Version 1.5N: A modification of HECTR Version 1.5 for application to N Reactor

International Nuclear Information System (INIS)

Camp, A.L.; Dingman, S.E.

1987-05-01

This report describes HECTR Version 1.5N, which is a special version of HECTR developed specifically for application to the N Reactor. HECTR is a fast-running, lumped-parameter containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. Version 1.5N is a modification of Version 1.5 and includes changes to the spray actuation logic, and models for steam vents, vacuum breakers, and building cross-vents. Thus, all of the key features of the N Reactor confinement can be modeled. HECTR is designed for flexibility and provides for user control of many important parameters, if built-in correlations and default values are not desired

Safety of 5 MW district heating reactor (DHR) and hydraulic dynamic pressure drive control rods

International Nuclear Information System (INIS)

Wu Yuanqiang; Wang Dazhong

1991-11-01

The principles and movement characteristic of the hydraulic dynamic pressure drive for control rods in 5 MW district heating reactor are described with stress on analysis of its effects on reactor safety features. The drive is different from electric-magnetic drive for PWR or hydraulic drive for BWR. The drive cylinder is driven by dynamic pressure. In the new drive system, the reactor coolant (water) used as actuating medium is pressed by pump, then injected into a step cylinder which is set in the reactor core. The cylinder will move step by step by controlling flow, then the cylinder drives the neutron absorber and controls nuclear reaction. The drive is characterized by simplicity in structure, high reliability, inherent safety, reduction in reactor height, economy, etc

Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

International Nuclear Information System (INIS)

Dybeck, P.; Gustafsson, B.

1990-01-01

This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

RELAP5 kinetics model development for the Advanced Test Reactor

International Nuclear Information System (INIS)

Judd, J.L.; Terry, W.K.

1990-01-01

A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs

A Study on consensus development through stake holder engagement for the use of nuclear energy in Japan

International Nuclear Information System (INIS)

Aoyama, Yoshiko; Sawada, Tetsuo; Fujii, Yasuhiko

2008-01-01

In order to alleviate the distrust surrounding the use of nuclear energy and operation of nuclear facilities, as well as to foster trust and find better ways to live in harmony with society, it is essential for the state and electric power utilities not to provide one-sided information, but to formulate adequate bilateral communication and dialogue with stake holders. In this study, we conducted a case study focused on four stake holder meetings with a view to ensuring safety in the use of nuclear energy and operation of nuclear facilities in Japan; the 'Community Meeting for Securing the Transparency of Kashiwazaki-Kariwa Nuclear Power Station (Chiiki-no-kai)'; the 'Monitoring and Evaluation Conference on Environmental Radiation', established by communities where nuclear facilities are installed; the 'Round-Table Conference on Nuclear Power Policy' and the 'Conference for Public Participation and Decision Making for Nuclear Energy Policy' organized by Japan Atomic Energy Commission. As a consequence, it has been observed that in each case varying opinions were summarized through stake holder meetings, but the fact that no system exists to reflect these opinions in design of the nuclear energy policy has emerged as a controversial point. This study suggests future efforts in regard to the requirements in Japan for stake holder meetings that are necessary for fostering trust and mutual understanding among stake holders including the state and local government, electric power utilities and local residents. (author)

Verification of surface contamination density standard using clearance automatic laser inspection system for objects from a nuclear power plant

International Nuclear Information System (INIS)

Sasaki, Michiya; Ogino, Haruyuki; Ichiji, Takeshi; Hattori, Takatoshi

2008-01-01

In the clearance level inspection in Japan, it is necessary to indicate that the activity level of the target object must be less than not only the clearance levels, but also the surface contamination density standards. The classification measurements for these two standards have been performed separately, and the GM survey meters based on beta-ray measurement have mainly been used for surface contamination density measurement so far. Recently the Clearance Automatic Laser Inspection System, named CLALIS, has been developed to estimate the low-level activity concentration. This system consists of 3-dimensional laser scanner for shape measurement and eight large NE102A plastic scintillation detectors for gamma-ray measurement, and it has been clarified that the CLALIS has adequate detection ability for clearance measurement of both metal scraps and concrete debris. In this study, we compared the surface contamination densities for a number of actual contaminated and non-contaminated objects generated inside from the radiation controlled area at the Kashiwazaki-Kariwa nuclear power station by using the CLALIS and the GM survey meter. As a result, since CLALIS could detect the surface contamination as well as the GM survey meter for all measurement targets, it was revealed that CLALIS can rationally achieve clearance level inspection in a single radiation measurement. The practicality of CLALIS in view of the detection limit and processing time was discussed by comparison with the usual radiation monitors for surface contamination measurement. (author)

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.

Operating reactors licensing actions summary. Volume 5, No. 2

International Nuclear Information System (INIS)

1985-04-01

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 Resource Management. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

Operating reactors licensing actions summary. Volume 5, No. 6

International Nuclear Information System (INIS)

1985-08-01

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 Resource Management. This summary report is published for internal NRC use in managing the Operating Reactors Licensing Actions Program. Its content will change based on NRC management informational requirements

Decolourization of remazol black-5 textile dyes using moving bed bio-film reactor

Science.gov (United States)

Pratiwi, R.; Notodarmojo, S.; Helmy, Q.

2018-01-01

The desizing and dyeing processes in the textile industries produces wastewaster containing high concentration of organic matter and colour, so it needs treatment before released to environment. In this research, removal of azo dye (Remazol Black 5/RB 5) and organic as COD was performed using Moving Bed Biofilm Reactor (MBBR). MBBR is biological treatment process with attached growth media system that can increase removal of organic matter in textile wastewater. The effectiveness of ozonation as pre-treatment process to increase the removal efficiency in MBBR was studied. The results showed that in MBBR batch system with detention time of 1 hour, pre-treatment with ozonation prior to MBBR process able to increase the colour removal efficiency of up to 86.74%. While on the reactor without ozone pre-treatment, the colour removal efficiency of up to 68.6% was achieved. From the continuous reactor experiments found that both colour and COD removal efficiency depends on time detention of RB-5 dyes in the system. The higher of detention time, the higher of colour and COD removal efficiency. It was found that optimum removal of colour and COD was achieved in 24 hour detention time with its efficiency of 96.9% and 89.13%, respectively.

RELAP5 - a new tool for pressurized water reactor safety analysis

International Nuclear Information System (INIS)

Perneczky, L.

1988-11-01

The RELAP type pressurized water reactor safety system codes are used world wide for the loss of coolant accident analyses. In this paper the RELAP5, the advanced generation of the code family is presented. The relationship to RELAP4/mod6 version is discussed. The capability of the RELAP5/mod1-EUR version for small, medium and large break LOCA is investigated based on international user experience. (author) 30 refs

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)

Operating reactors licensing actions summary. Volume 5, No. 7

International Nuclear Information System (INIS)

1985-09-01

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 Resource Management

Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

International Nuclear Information System (INIS)

Rimpault, G.

2004-01-01

Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

RELAP5 analyses of overcooling transients in a pressurized water reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.; Ogden, D.M.; Stitt, B.D.; Waterman, M.E.

1983-01-01

In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.5 computer code. These analyses were performed for Oconee Plants 1 and 3, which are pressurized water reactors of Babcock and Wilcox lowered-loop design. Results of the RELAP5 analyses are presented, including a comparison with plant data. The capabilities and limitations of the RELAP5/MOD1.5 computer code in analyzing integral plant transients are examined. These analyses require detailed thermal-hydraulic and control system computer models

An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

International Nuclear Information System (INIS)

Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

1976-01-01

It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

Emergency planning and emergency drill for a 5 MW district heating reactor

International Nuclear Information System (INIS)

Shi Zhongqi; Wu Zhongwang; Hu Jingzhong; Feng Yuying; Li Zhongsan; Dong Shiyuan

1991-01-01

The authors describes the main contents of the emergency planning for a 5 MW nuclear district heating reactor and some considerations for the planning's making, and presents the situation on implementing emergency preparedness and an emergency drill that has been carried out

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Directory of Open Access Journals (Sweden)

Skrzypek Maciej

2015-09-01

Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

Development of Advanced Concept for Shortening Construction Period of ABWR Plant

International Nuclear Information System (INIS)

Hiroshi Ijichi; Toshio Yamashita; Masahiro Tsutagawa; Hiroya Mori; Nobuaki Ooshima; Jun Miura; Minoru Kanechika; Nobuaki Miura

2002-01-01

Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO's next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipment installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts.) In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment. (authors)

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

Science.gov (United States)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis

International Nuclear Information System (INIS)

Hoogenboom, J.E.

2013-01-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)

WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

International Nuclear Information System (INIS)

2004-01-01

1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

Operating reactors licensing actions summary. Volume 5, Number 1

International Nuclear Information System (INIS)

1985-03-01

This document 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 Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

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

Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

1978-12-01

This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

Maintenance of Structures, Systems and Components of the RSG-GAS Reactor as an Implementation of BAPETEN Regulation No. 5 Year 2011

International Nuclear Information System (INIS)

Aep Saepudin Catur; Dede Solehudin Fauzi; Djunaidi

2012-01-01

Maintenance activities for structures, systems and components of the reactor, consider prerequisites for the operation of the non power reactor. This activity is intended to ensure that the structures, systems and components function properly. Implementation of reactor maintenance carried out starting from programs establishment, scheduling, maintenance accomplishment and maintenance report. This paper will describe the implementation of reactor maintenance non power as required by BAPETEN regulation no.5, year 2011. By understanding correctly of this regulation, it is expected that maintenance activity of structures, systems and components of the reactor can be successfully performed. The RSG-GAS reactor has implemented various types of reactor maintenance based on BAPETEN regulation no.5 year 2011 properly. As a result failure of the structures, systems and components of the reactor can be minimized then they can be kept reliable. (author)

On results of measurement and method of behavior analysis for land slide protection wall in excavation works for main building foundation of No.2 plant in Kashiwazaki-Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc

International Nuclear Information System (INIS)

Otsuki, Hideo; Tsuchi, Hiroyuki; Nishioka, Toshimichi

1987-01-01

Tokyo Electric Power Co. has been constructing the nuclear power station having 8 million kW capacity of seven BWR plants in the site of about 4.2 million m 2 in Niigata Prefecture. No.1 BWR plant of 1100 MWe output started the operation in September, 1985. As a rule, the important structures in nuclear power stations such as a reactor building and a turbine building are to be directly supported on bedrocks, and in this case, on the mudstone of Nishiyama strata. As this Nishiyama strata exists in large depth, the excavation works for the foundations of buildings are to be carried out by installing large scale land slide protection walls. In this report, among the excavation works for the main building foundation of No.2 plant, the results of examining the behavior of the land slide protection wall installed in soft rock ground based on the results of measurement of vertical excavation by land slide protection method and the techniques of its analysis are described. The geological features, the design of land slide protection walls, the measurement of the land slide protection walls and surrounding ground and the results, and the examination of the analysis methods by a beam model and FEM are reported. (Kako, I.)

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)

High Flux Isotope Reactor system RELAP5 input model

International Nuclear Information System (INIS)

Morris, D.G.; Wendel, M.W.

1993-01-01

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model

On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

International Nuclear Information System (INIS)

Causey, A.R.; Holt, R.A.

1993-06-01

Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

International Nuclear Information System (INIS)

Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

1995-01-01

This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs

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)

Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

Energy Technology Data Exchange (ETDEWEB)

Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2015-09-15

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Operating reactors licensing actions summary. Volume 5, No. 9

International Nuclear Information System (INIS)

1985-11-01

This document 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 Resource Management

Operating reactors licensing actions summary. Volume 5, No. 8

International Nuclear Information System (INIS)

1985-10-01

This 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 Resource Management

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.

Research activities on high-temperature gas-cooled reactors (HTRs) in the 5. EURATOM RTD Framework programme

International Nuclear Information System (INIS)

Martin-Bermejo, J.; Hugon, M.; Van Goethem, G.

2002-01-01

One of the areas of research of the 'nuclear fission' key action of the 5. EURATOM RTD Framework Programme (FP5) is the safety and efficiency of future systems. The main objective of this area is to investigate and evaluate new or revisited concepts (both reactors and alternative fuels) for nuclear energy that offer potential longer term benefits in terms of cost, safety, waste management, use of fissile material, less risk of diversion and sustainability. Several projects related to high-temperature gas-cooled reactors (HTRs) were retained by the European Commission (EC) services. They address important issues such as HTR fuel technology, HTR fuel cycle, HTR materials, power conversion systems and licensing. Most of these projects have already started and are progressing according to the schedule. They are the initial core of activities of a European Network on 'High-temperature Reactor Technology' (HTR-TN) recently set up by 18 EU organisations. (authors)

Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

International Nuclear Information System (INIS)

Bhattacharyya, S. K.; Boing, L. E.

2000-01-01

The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

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

1989-01-01

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

Research reactor 'A' 6.5/10 MW; Istrazivacki reaktor 'A' 6,5/10 MW

Energy Technology Data Exchange (ETDEWEB)

Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1961-07-01

This booklet includes a short description of the RA research reactor with basic properties of its components: control and safety system, heavy water system, technical water cooling system, heavy water distillation system, cover gas system, dosimetry control system, power supply system. It is used for fundamental research in reactor and nuclear physics, isotope production, materials testing.

Neutronics modeling of TRIGA reactor at the University of Utah using agent, KENO6 and MCNP5 codes

International Nuclear Information System (INIS)

Yang, X.; Xiao, S.; Choe, D.; Jevremovic, T.

2010-01-01

The TRIGA reactor at the University of Utah is modelled in 2D using the AGENT state-of-the-art methodology based on the Method of Characteristics (MOC) and R-function theory supporting detailed reactor analysis of reactor geometries of any type. The TRIGA reactor is also modelled using KENO6 and MCNP5 for comparison. The spatial flux and reaction rates distribution are visualized by AGENT graphics support. All methodologies are in use in to study the effect of different fuel configurations in developing practical educational exercises for students studying reactor physics. At the University of Utah we train graduate and undergraduate students in obtaining the Nuclear Regulatory Commission license in operating the TRIGA reactor. The computational models as developed are in support of these extensive training classes and in helping students visualize the reactor core characteristics in regard to neutron transport under various operational conditions. Additionally, the TRIGA reactor is under the consideration for power uprate; this fleet of computational tools once benchmarked against real measurements will provide us with validated 3D simulation models for simulating operating conditions of TRIGA. (author)

Development of a coupled reactor with a catalytic combustor and steam reformer for a 5 kW solid oxide fuel cell system

International Nuclear Information System (INIS)

Kang, Sanggyu; Lee, Kanghun; Yu, Sangseok; Lee, Sang Min; Ahn, Kook-Young

2014-01-01

Highlights: • Proposes the scale-up strategy to develop a large-scale coupled reactor. • Investigation of performance of steam reformer coupled with catalytic combustor. • Experimental parameters are inlet temp., air excess ratio, SCR, fuel utilization. • Evaluation of the heat transfer distribution along the gas flow direction. • The mean value of methane conversion rate is approximately 93.4%. - Abstract: The methane (CH 4 ) conversion rate of a steam reformer can be increased by thermal integration with a catalytic combustor, called a coupled reactor. In the present study, a 5 kW coupled reactor has been developed based on a 1 kW coupled reactor in previous work. The geometric parameters of the space velocity, diameter and length of the coupled reactor selected from the 1 kW coupled reactor are tuned and applied to the design of the 5 kW coupled reactor. To confirm the scale-up strategy, the performance of 5 kW coupled reactor is experimentally investigated with variations of operating parameters such as the fuel utilization in the solid oxide fuel cell (SOFC) stack, the inlet temperature of the catalytic combustor, the excess air ratio of the catalytic combustor, and the steam to carbon ratio (SCR) in the steam reformer. The temperature distributions of coupled reactors are measured along the gas flow direction. The gas composition at the steam reformer outlet is measured to find the CH 4 conversion rate of the coupled reactor. The maximum value of the CH 4 conversion rate is approximately 93.4%, which means the proposed scale-up strategy can be utilized to develop a large-scale coupled reactor

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.)

In-reactor creep of zirconium-2.5 wt% niobium at 570 K

International Nuclear Information System (INIS)

Coleman, C.E.; Causey, A.R.; Fidleris, V.

1976-01-01

The effect of fast neutron flux at 570 K on the creep rate of specimens of zirconium-2.5 wt% niobium alloy taken from tubes in various metallurgical conditions has been measured using both constant load tensile creep machines and bent-beam stress relaxation. Creep rates calculated from stress relaxation fit on the trend line for the constant load creep data. Between 114 MPa and 450 MPa the creep rate is proportional to neutron flux. The creep rate of specimens from the longitudinal direction is about twice that of specimens from the circumferential direction of a tube. This anisotropy in creep strength is attributed partly to crystallographic texture and partly to deformation substructure. Cold-work is detrimental to in-reactor creep strength; as-extruded material has higher creep strength. In cold-worked material at stresses below 100 MPa the stress exponent, n, is about 1; n gradually increases with stress being about 10 at 525 MPa and about 100 at 660 MPa. In laboratory tests, rupture ductility correlates inversely with n; the lower n the higher the ductility. In-reactor tests support this correlation thus pressure tubes in CANDU reactors, operating at 117 MPa where n approximately 1, should have good ductility. (Auth.)

Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

International Nuclear Information System (INIS)

Reis, Patrícia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M.; Veloso, Maria Auxiliadora F.

2012-01-01

Highlights: ► We investigated how much the code results are affected by the code user. ► Two essential modifications were made on a previously validated nodalization. ► We used the RELAP5 code to predict the results. ► Results highlight the necessity of sensitivity analysis to have the ideal modeling. - Abstract: The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.

Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

Directory of Open Access Journals (Sweden)

Noble Brooklyn

2012-01-01

Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.

2007-01-01

Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

Degradation of Reactive Black 5 dye using anaerobic/aerobic membrane bioreactor (MBR) and photochemical membrane reactor

International Nuclear Information System (INIS)

You, Sheng-Jie; Damodar, Rahul A.; Hou, Sheng-Chon

2010-01-01

Three different types of advance treatment methods were evaluated for the degradation of Reactive Black 5 (RB5). The performance of two stage anaerobic SBR-aerobic MBR, anaerobic MBR with immobilized and suspended biocells and an integrated membrane photocatalytic reactor (MPR) using slurry UV/TiO 2 system were investigated. The results suggest that, nearly 99.9% color removal and 80-95% organic COD and TOC removal can be achieved using different reactor systems. Considering the Taiwan EPA effluent standard discharge criteria for COD/TOC, the degree of treatment achieved by combining the anaerobic-aerobic system was found to be acceptable. Anew, Bacilluscereus, high color removal bacterium was isolated from Anaerobic SBR. Furthermore, when this immobilized into PVA-calcium alginate pellets, and suspended in the anaerobic MBR was able to achieve high removal efficiencies, similar to the suspended biocells system. However, the immobilized cell Anaerobic MBR was found to be more advantageous, due to lower fouling rates in the membrane unit. Results from slurry type MPR system showed that this system was capable of mineralizing RB5 dyes with faster degradation rate as compared to other systems. The reactor was also able to separate the catalyst effectively and perform efficiently without much loss of catalyst activity.

Description of the two-loop RELAP5 model of the L-Reactor at the Savannah River Site

International Nuclear Information System (INIS)

Cozzuol, J.M.; Davis, C.B.

1989-12-01

A two-loop RELAP5 input model of the L-Reactor at the Savannah River Site (SRS) was developed to support thermal-hydraulic analysis of SRS reactors. The model was developed to economically evaluate potential design changes. The primary simplifications in the model were in the number of loops and the detail in the moderator tank. The six loops in the reactor were modeled with two loops, one representing a single loop and the other representing five combined loops. The model has undergone a quality assurance review. This report describes the two-loop model, its limitations, and quality assurance. 29 refs., 18 figs., 10 tabs

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

International Nuclear Information System (INIS)

Szczurek, J.

1995-01-01

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J. [Inst. of Atomic Energy, Swierk (Poland)

1995-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

Energy Technology Data Exchange (ETDEWEB)

Szczurek, J [Inst. of Atomic Energy, Swierk (Poland)

1996-12-31

The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

RELAP5 thermal-hydraulic analyses of overcooling sequences in a pressurized water reactor

International Nuclear Information System (INIS)

Bolander, M.A.; Fletcher, C.D.; Davis, C.B.; Kullberg, C.M.; Stitt, B.D.; Waterman, M.E.; Burtt, J.D.

1984-01-01

In support of the Pressurized Thermal Shock Integration Study, sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.6 and MOD2.0 computer codes. These analyses were performed for the H.B. Robinson Unit 2 pressurized water reactor, which is a Westinghouse 3-loop design plant. Results of the RELAP5 analyses are presented. The capabilities of the RELAP5 computer code as a tool for analyzing integral plant transients requiring a detailed plant model, including complex trip logic and major control systems, are examined

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Integrated analysis for a small break LOCA in the IRIS reactor using MELCOR and RELAP5 codes

International Nuclear Information System (INIS)

Del Nevo, A.; Manfredini, A.; Oriolo, F.; Paci, S.; Oriani, L.

2004-01-01

The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002 and IRIS is one of the designs considered by US utilities as part of the ESP (Early Site Permit) process. This paper's focus is on the use of well known computer codes for integrated (reactor vessel and containment) calculations of the IRIS response to a small break loss of coolant accident (LOCA). In IRIS, large break LOCA events are eliminated by the use of a layout configuration in which the reactor vessel contains all the reactor coolant system components including the core, control rod drive mechanisms, pressurizer, steam generators, and coolant pumps. Thus the IRIS configuration has no large loop piping; also, no pipes with a diameter greater than 0.1 meters are part of the reactor coolant system boundary. For small break LOCAs, IRIS features an innovative mitigation approach that is based on maintaining coolant inventory rather than designing high and low pressure injection systems to provide makeup coolant to the reactor to maintain core cooling. The novel IRIS approach requires development of evaluation models that are different from those used for the current generation of pressurized water reactors. An analysis of small break LOCAs for IRIS is documented in two companion papers, and has been developed using a preliminary evaluation model based on the explicit coupling of the RELAP5 and GOTHIC codes. The objective of this paper is to compare the results obtained via the coupled RELAP/GOTHIC code with different computational tools. A reference case from the preliminary IRIS safety assessment was selected, and the same small break LOCA sequence is analyzed using

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

RA research nuclear reactor operation in forced regime, Annex 5; Prilog 5 - Rad istrazivackog nuklearnog reaktora RA u forsiranom rezimu, Prvo saopstenje

Energy Technology Data Exchange (ETDEWEB)

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

1964-12-15

In order to increase the flux and experimental potentials of the RA reactor, properties of the reactor operating in the forced regime have been inspected four times, twice with the spent fuel (core C), and twice with partly fresh fuel (cores E and F). Forced regime means power higher than nominal 6.5 MW. From the analyses of technology parameters during operation in the forced regime at power levels of 7 and 11 MW, state and behaviour of the components, efficiency of the biological protection and other parameters it was concluded that the reactor operation is possible at power levels of 10 MW during a whole year, and under certain conditions even longer operation at power levels up to 15 MW. This paper examines the reactor operation conditions in the forced regime and gives conclusions related to real increase of the experimental possibilities, isotope production and economic factors. Radi povecanja fluksa i eksperimentalnih mogucnosti reaktora RA, cetiri puta su proverene osobine reaktora u forsiranom rezimu, tj. na snagama vecim od nominalne (6,5 MW) i to dva puta sa istrosenim gorivom (jezgro C) i dva puta sa delimicno svezim (jezgra E i F). Iz analize tehnoloskih parametara u toku eksperimentalno rada reaktora u forsiranom rezimu na snagama od 7 do 11 MW, stanja i ponasanja opreme, efikasnosti bioloske zastite i ostalih parametara, zakljucuje se da je moguc permanentan rad reaktora tokom cele godine na snagama do 10 MW, a pod izvesnim uslovima i duzi rad na snagama do 15 MW. U radu su diskutovani uslovi rada reaktora u forsiranom rezimu i donose se zakljucci o realnom povecanju eksperimentalnih mogucnosti, proizvodnje radioizotopa i o ekonomskim faktorima (author)

CFX-10 and RELAP5-3D simulations of coolant mixing phenomena in RPV of VVER-1000 reactors

International Nuclear Information System (INIS)

Terzuoli, F.; Moretti, F.; Melideo, D.; D'Auria, F.; Shkarupa, O.

2006-01-01

The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed with the ANSYS CFX-10 CFD code and with the RELAP5-3D system code. In particular, the attention focused on the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. The results have been compared against experimental data from V1000CT-2 Benchmark. (author)

Thermal, thermo-hydraulic and thermo-mechanic analysis for fuel elements of IEA-R1 reactor at 5MW

International Nuclear Information System (INIS)

Teixeira e Silva, A.; Silva Macedo, L.V. da

1989-01-01

In connection with the on going conversion of IEA-R1 Research Reactor, operated by IPEN-CNEN/SP, from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel, steady-state thermal and thermo-hydraulic analysis of both existing HEU and proposed LEU cores under 2 MW operating conditions have been carried out. Keeping in mind the possibility of power upgrading, steady-state thermal, thermo-hydraulic and thermomechanical analysis of proposed LEU core under 5 MW operating conditions have also been carried out. The thermal and thermo-hydraulic analysis at 2 MW show that the conversion of the existing HEU core to be proposed LEU core will not change the reactor safety margins. Although the upgrading of the reactor power to 5 MW will result in safety margins lower than in case of 2MW, these will be still sufficient for optimum operation and safe behaviour. The thermomechanical analysis at 5 MW show that the thermal stresses induced in the fuel element will satisfy the design limits for mechanical strenght and elastic stability. (author) [pt

Development and validation of a model TRIGA Mark III reactor with code MCNP5

International Nuclear Information System (INIS)

Galicia A, J.; Francois L, J. L.; Aguilar H, F.

2015-09-01

The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

International Nuclear Information System (INIS)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Aronne, Ivan D.; Rezende, Guilherme P.

2011-01-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

Analysis of loss of flow events on Brazilian multipurpose reactor by RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Soares, Humberto V.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria Auxiliadora F., E-mail: antonella@nuclear.ufmg.br, E-mail: laubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, UFMG, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores, CNPq (Brazil); Aronne, Ivan D.; Rezende, Guilherme P., E-mail: aroneid@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil).

2011-07-01

The Brazilian Multipurpose Reactor (BMR) is currently being projected and analyzed. It will be a 30 MW open pool multipurpose research reactor with a compact core using Materials Testing Reactor (MTR) type fuel assembly, with planar plates. BMR will be cooled by light water and moderated by beryllium and heavy water. This work presents the calculations of steady state operation of BMR using the RELAP5 model and also three transient cases of loss of flow accident (LOFA), in the primary cooling system. A LOFA may arise through failures associated with the primary cooling system pumps or through events resulting in a decrease in the primary coolant flow with the primary cooling system pumps functioning normally. The cases presented in this paper are: primary cooling system pump shaft seizure, failure of one primary cooling system pump motor and failure of both primary cooling system pump motors. In the shaft seizure case, the flow reduction is sudden, with the blocking of the flow coast down The motor failure cases, deal with the failure of one or two pump motor due to, for example, malfunction or interruption of power and differently of the shaft seizure it can be observed the flow coast down provided by the pump inertia. It is shown that after all initiating events the reactor reaches a safe new steady state keeping the integrity of the fuel elements. (author)

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

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

Volatility of V15Cr5Ti fusion reactor alloy

International Nuclear Information System (INIS)

Neilson, R.M. Jr.

1986-01-01

One potential hazard from the presence of activation products in fusion facilities is accidental oxidation-driven volatility of those activation products. Scoping experiments were conducted to investigate the oxidation and elemental volatility of candidate fusion reactor alloy V15Cr5Ti as a function of time, temperature, and test atmosphere. Experiments in air and in argon carrier gases containing 10 4 to 10 1 Pa (10 -1 to 10 -4 atm) oxygen were conducted to investigate the lower oxygen partial pressure limit for the formation of a low melting point (approximately 650 0 C), high volatility, oxide layer and its formation rate. Experiments to determine the elemental volatility of alloy constituents in air at temperatures of 700 0 C to greater than 1600 0 C. Some of these volatility experiments used V15Cr5Ti that was arc-remelted to incorporate small quantities (<0.1 wt. %) of Sc and Ca. Incorporation of Sc and Ca in test specimens permitted volatility measurement of radioactive constituents present only after activation of V15Cr5Ti

Operation status display and monitoring system for BWR nuclear power plant

International Nuclear Information System (INIS)

Wakabayashi, Yasuo; Hayakawa, Hiroyasu; Kawamura, Atsuo; Kaneda, Mitsunori.

1982-01-01

Lately, the development of the system has been made for BWR plants, which monitors the operating status not only in normal operation but also in abnormal state and also for plant safety. Recently, the improvement of man-machine interface has been tried through the practical use of technique which displays data collectively on a CRT screen relating them mutually. As one of those results, the practical use of an electronic computer and color CRT display for No. 1 unit in the Fukushima No. 2 Nuclear Power Station (2F-1), Tokyo Electric Power Co., is described. Also, new centralized control panels containing such systems were used for the 1100 MWe BWR nuclear power plants now under construction, No. 3 unit of the Fukushima No. 2 Power Station and No. 1 unit of Kashiwazaki-Kariwa Nuclear Power Station (2F-3 and K-1, respectively). The display and monitoring system in 2F-1 plant is the first one in which a computer and color CRTs were practically employed for a BWR plant in Japan, and already in commercial operation. The advanced operating status monitoring system, to which the result of evaluation of the above system was added, was incorporated in the new centralized control panels presently under production for 2F-3 and K-1 plants. The outline of the system, the functions of an electronic computer, plant operating status monitor, surveillance test guide, the automation of plant operation and auxiliary operation guide are reported for these advanced monitoring system. It was confirmed that these systems are useful means to improve the man-machine communication for plant operation minitoring. (Wakatsuki, Y.)

The alkaline pre-filming process of the RWCU piping surface

Energy Technology Data Exchange (ETDEWEB)

Kao, D-Y.; Wen, T-J., E-mail: dahyukao@iner.gov.tw [Atomic Energy Council, Inst. of Nuclear Energy Research, Lungtan Taoyuan, Taiwan (China); Fong, C., E-mail: clinton@itri.org.tw [Industrial Technology Research Inst., Chutung, Hsinchu (China); Lu, J-H., E-mail: u879864@taipower.com.tw [Taiwan Power Co., Lungmen Nuclear Power Plant, Gongliao Township, Taipei County (China)

2010-07-01

The radiation dose rate on a BWR can cause great damage to operators and their maintenances working quality, so it is very important to lower it. The RWCU of a BWR primary system has the highest radiation dose rate and deposition of an active corrosion product because of: continually cleaning the reactor water and the repeated, acute change of temperature. The Lungmen NPP was built as an advanced BWR (ABWR) system. All new ABWRs in Japan, such as Kashiwazaki-Kariwa NPP units K6 and K7, Hamaoka NPP unit H5, and Shika NPP unit2, were given in passivation treatment in the stage of nuclear heating or flow-induced vibration (FIV) testing before each plant's initial start up. The conditions of alkaline pre-filming process include: pH, dissolve oxygen (DO) and temperature. This project had been decided by the results of earlier water loop testing. Three kinds of specimens including stainless steel AISI SS316L and carbon steel SA106 GrB, SA333 Gr6 were put into the autoclave located at a bypass system during the process of passivation treatment. The materials of these specimens have been used in RWCU piping, and during the alkaline pre-filming process they were monitored by AC impedances and electrochemical corrosion potential (ECP). Their surface analyses including SEM, EDS, ESCA and SIMS data were obtained from the initial test, and obtained every six months thereafter. From the SEM plot of first issue it was found that the specimens' surface of all three different materials presented excellent levels of oxidized pellets. They are almost completely packed, for both large and small granules, and the porosity is less than 40%. The deposition tests of cobalt ion were also completed. The material of stainless steel was nearly not affected through the passive process. However, carbon steel could reduce about 30% to 50% of deposited cobalt ion. This result follows the conclusion of Hatch Company. Based on the nominal cost of radiation protection, i.e. 10000 USD

Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

International Nuclear Information System (INIS)

Bottimore, R.R.

1980-12-01

The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

A preliminary neutronic evaluation and depletion study of VHTR and LS-VHTR reactors using the codes: WIMSD5 and MCNPX

International Nuclear Information System (INIS)

Silva, Fabiano C.; Pereira, Claubia; Costa, Antonella Lombardi; Veloso, Maria Auxiliadora Fortini

2009-01-01

It is expected that, in the future, besides electricity generation, reactors should also develop secondary activities, such as hydrogen generation and seawater desalinization. Generation IV reactors are expected to possess special characteristics, like high safety, minimization of radioactive rejects amount and ability to use reprocessed fuel with non-proliferating projects in their cycles. Among the projects of IV generation reactors available nowadays, the (High Temperature Reactors) HTR, are highlighted due to these desirable characteristics. Under such circumstances, such reactor may be able to have significant higher thermal power ratings to be used for hydrogen production, without loose of safety, even in an emergency. For this work, we have chosen two HTR concepts of a prismatic reactor: (Very High Temperature Reactor) VHTR and the (Liquid Salted -Very High Temperature Reactor) LS-VHTR. The principal difference between them is the coolant. The VHTR uses helium gas as a coolant and have a burnup of 101,661 MWd/THM while the LS-VHTR uses low-pressure liquid coolant molten fluoride salt with a boiling point near 1500 de C working at 155,946 MWd/THM. The ultimate power output is limited by the capacity of the passive decay system; this capacity is limited by the reactor vessel temperature. The goal was to evaluate the neutronic behavior and fuel composition during the burnup using the codes (Winfrith Improved Multi-Group Scheme) WIMSD5 and the MCNPX2.6. The first, deterministic and the second, stochastic. For both reactors, burned fuel type 'C' coming from Angra-I nuclear plant, in Brazil, was used with 3.1% of initial enrichment, burnup to 33,000 MWd/THM using the ORIGEN2.1 code, divided in three steps of 11,000 MWd/THM, with an average density power of 37.75 MWd/THM and 5 years of cooling in pool. Finally, the fuel was reprocessed by Purex technique extracting 99.9% of Pu, and the desired amount of fissile material (15%) to achieve the final mixed oxide was

Japan country report

Energy Technology Data Exchange (ETDEWEB)

Morisaki, Rieko [Energy Communication Planning, 3-9-16 Aobadai, 818-0137 Dazaifu (Japan)

2008-07-01

1. Nuclear 2007 highlights: - A magnitude 6.8 earthquake occurred in Niigata on July 16 2007. Owing to this earthquake, 3 units operating and 1 unit during start-up were shutdown automatically at TEPCO's Kashiwazaki-Kariwa NPS. Now, all 7 units of the NPS are in an outage for investigation. This influenced the capacity factor of Japanese NPPs in FY 2007, which stood at just 60.7%. - Debate on global warming is more and more active in Japan, as it is the host country of the G8 Hokkaido Toyako Summit in July. The Japan Atomic Energy Commission released 'White Paper on Nuclear Energy 2007' in March 2008. In the paper, they first expressed the view that the expansion of the peaceful use of nuclear energy is indispensable. 2. Nuclear overview: a. Energy policy: Electricity share: 25.4% of nuclear. The energy policy of Japan aims at nuclear power generation being maintained at the current level (30 to 40% of the total electricity generation) or increasing even after 2030, for stable energy supply and as a countermeasure against global warming. - Nuclear Fuel Cycle: The active tests at the JNFL reprocessing plant in Rokkasho-mura are in the final phase for commercial operation in 2008. By FY 2010 Plutonium utilization in LWRs in 16 to 18 NPP units. Around FY 2010 Installment of new centrifuges at the uranium enrichment plant at Rokkasho-mura. In FY 2012 Start of commercial operation of MOX fuel fabrication plant. Fast-breeder reactor cycle: Operation of the prototype reactor 'MONJU' has been suspended since a secondary sodium leak in 1995. JAEA (Japan Atomic Energy Agency) completed full-scale remodeling work and is implementing various tests to confirm the capabilities and soundness of MONJU. They aim to start its operation within FY 2008. Around 2025 Building a demonstration FBR. Before 2050 Development of a commercial FBR. - Electricity production (Operating): BWR: 32 units (including 4 units of APWR), PWR: 23 units. - Electricity production

Probability safety assessment of the Kozloduy-5 and Kozloduy-6 reactors

Energy Technology Data Exchange (ETDEWEB)

Boyadzhiev, A; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

A probability safety assessment (PSA) of Level 1 (assessment of plant failures leading to the determination of core damage frequency) has been carried out for the NPP Kozloduy Units 5 and 6 (reactors WWER-1000). The scope of the study includes all significant accident initiators including seismic (earthquake) and fire initiators. Event trees for all initiators and fault trees for front line systems, support systems and major safety systems have been built. A distribution of the different initiators has been established as follows: internal initiators - 85%, seismic initiators - 5%, fire initiators- 10%. The loss of offsite power was identified as main contributor from the internal initiators with frequency 1,1.10{sup -4}/y. It is concluded that the safety functions of WWER-1000 are adequately covered by the safety systems. 4 refs., 2 tabs.

Application of non-thermal plasma reactor for degradation and detoxification of high concentrations of dye Reactive Black 5 in water

Directory of Open Access Journals (Sweden)

Dojčinović Biljana P.

2016-01-01

Full Text Available Degradation and detoxification efficiency of high concentrations of commercially available reactive textile dye Reactive Black 5 solution (40, 80, 200, 500, 1000 mg L-1, were studied. Advanced oxidation processes in water falling film based dielectric barrier discharge as a non-thermal plasma reactor were used. For the first time, this reactor was used for the treatment of high concentrations of organic pollutants such as reactive textile dye Reactive Black 5 in water. Solution of the dye is treated by plasma as thin water solution film that is constantly regenerated. Basically, the reactor works as a continuous flow reactor and the electrical discharge itself takes place at the gas-liquid interphase. The dye solution was recirculated through the reactor with an applied energy density of 0-374 kJ L-1. Decolorization efficiency (% was monitored by UV-VIS spectrophotometric technique. Samples were taken after every recirculation (~ 22 kJ L-1 and decolorization percent was measured after 5 min and 24 h of plasma treatment. The efficiency of degradation (i.e. mineralization and possible degradation products were also tracked by determination of the chemical oxygen demand (COD and by ion chromatography (IC. Initial toxicity and toxicity of solutions after the treatment were studied with Artemia salina test organisms. Efficiency of decolorization decreased with the increase of the dye concentration. Complete decolorization, high mineralization and non-toxicity of the solution (<10 % were acomplished after plasma treatment using energy density of 242 kJ L-1, while the initial concentrations of Reactive Black 5 were 40 and 80 mg L-1. [Projekat Ministarstva nauke Republike Srbije, br. 172030 i br. 171034

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

5. symposium on nuclear reactor remote monitoring

International Nuclear Information System (INIS)

1987-01-01

17 papers deal with the data-technological concept and mode of operation of nuclear-reactor remote-monitoring (RM) systems from the perspectives of users in Baden-Wurttemberg, Sleswig-Holstein, Bavaria and Belgium, with the requirements on measuring devices and equipment in NRM systems, computer-controlled evaluation and processing of measured data, in particular the LASAT and OLDES systems. (DG) [de

Radiographic inspection and densitometric evaluation of CP-5 reactor fuel

International Nuclear Information System (INIS)

Staroba, J.F.; Knoerzer, T.W.

1978-02-01

This report covers the radiographic and densitometric techniques used as part of a quality verification program for CP-5 reactor fuel by the Nondestructive Assay Section of the Special Materials Division. Other nondestructive tests used were ultrasonic and gamma-ray spectrometry. The main objectives were to perform a one-hundred percent radiographic inspection of the fuel tubes and to derive a quantitative relationship between fuel thickness and film density with the use of fabricated fuel step wedges. By the use of tangential x-ray techniques, measurements were made of fuel peaks or ''hot spots'' that protruded above the main fuel line. Other general problems in radiographic inspection and solutions for the upgrading of the total radiographic inspection program are also discussed

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

Actualización del sistema SCADA y de control para los reactores MQ5 y MQ6 de la planta de Pinturas Condor, Sherwin Williams Ecuador

Directory of Open Access Journals (Sweden)

Jonathan Reinoso

2013-12-01

Full Text Available El presente documento describe la actualización del sistema SCADA para los reactores MQ5 y MQ6 de la planta de Pinturas Condor mediante el software Intouch y la actualización del sistema de control del reactor MQ5 implementado en un controlador lógico programable (PLC de marca SCHNEIDER, además de la arquitectura de control realizada en el proyecto. El sistema SCADA y de control de los reactores permiten la visualización y control de los datos y variables más relevantes durante las diferentes fases de producción de resinas en los reactores MQ5 y MQ6.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

Nuclear cycle of thorium and molt salts reactors. PE 5.8

International Nuclear Information System (INIS)

Doubre, H.

2004-01-01

In the framework of the nuclear industry development, many scenario are studied from the standard reactors using enriched uranium to the IV generation reactors. The study of new systems for the future of the nuclear needs to develop new simulation tools. The research programs of the IPN of Orsay are presented. (A.L.B.)

Implementing and measuring safety goals and safety culture. 4. Utility's Activities for Better Safety Culture After the JCO Accident

International Nuclear Information System (INIS)

Omoto, Akira

2001-01-01

The criticality accident at the JCO plant prompted the Government to enact a law for nuclear emergency preparedness. The nuclear industry established NSnet to facilitate opportunities for peer review among its members. This paper describes the activities by NSnet and TEPCO's Kashiwazaki-Kariwa nuclear power station (NPS) for a better safety culture. Created as a voluntary organization by the nuclear industry in 1999, NSnet has 35 members and is assisted by CRIEPI and NUPEC for its activities relevant to human factors. Given the fact that nuclear facility operators not belonging to WANO had no institutional system available for exchange of experiences and good practices for better safety among themselves, NSnet's activities focus on peer review by member organizations and onsite seminars. Starting April 2000 with visits to three fuel fabricators, NSnet intends to have 23 peer-review visits in 2 yr (Ref. 1). The six-member review team stays on-site for 4 days, during which time they review-using guidelines available from WANO and IAEA-OSART-six areas: organization/management, emergency preparedness, education/training, operation/ maintenance, protection against occupational radiation exposure, and prevention of accidents. A series of on-site seminars is held at members' nuclear facilities, to which NSnet dispatches experts for lectures. NSnet plans to hold such seminars twice per month. Other activities include information-sharing through a newsletter, a Web site (www. nsnet.gr.jp), and others. Although considerable differences exist in the design and the practices in operation/maintenance between power reactors and JCO, utilities can extract lessons from the accident that will be worth consideration for their own facilities in the areas of safety culture, education and training, and interface between design and operation. This thinking prompted the Nuclear Safety Promotion Center at Kashiwazaki-Kariwa NPS, to which the author belonged at that time, to launch the

TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

Science.gov (United States)

Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

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

Reactor lattice codes

International Nuclear Information System (INIS)

Kulikowska, T.

2001-01-01

The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

Computerized reactor power regulation with logarithmic controller

International Nuclear Information System (INIS)

Gossanyi, A.; Vegh, E.

1982-11-01

A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

The Simulator Development for RDE Reactor

Science.gov (United States)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

2016-04-15

Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

Recommendations of the Reactor Safety Commission (RSK) 1981-1983. Vol. 5

International Nuclear Information System (INIS)

1984-02-01

Since the reorganization of the RSK in 1971 the recommendations are published in the ''Bundesanzeiger'' (BAnz). This makes the work of the RSK transparent for the public. After publication of the recommendations in the Bundesanzeiger, the Office of the RSK publishes them once more as a summary report by order of the Federal Minister of Interior. The reports are divided generally into two parts: Part I contains the recommendations which are given by the RSK; part II contains the official notices concerning the RSK. There is also a subject index. In addition volume 5 contains as part III all papers which were held on the occasion of the 25th anniversary of the Reactor Safety Commission. (orig./HP) [de

Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5; Calculo de activacion neutronica de barras de control de un reactor nuclear, utilizando MCNP5

Energy Technology Data Exchange (ETDEWEB)

Pena V, J.D.

2016-07-01

The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Dosimetry and radiation shielding at the RA reactor, Annual report 1975, Annex 5

International Nuclear Information System (INIS)

Ninkovic, M.

1976-01-01

In the working environment at the RA reactor, the level of gamma radiation is measured continuously by the built-in stationary system. According to the needs, measurement are done in the reactor hall every day. The level of gamma radiation is measured separately in typical points when the reactor is operated at nominal power and during intervals between two operating campaigns. The level of neutron radiation is measured according to the needs by means of a mobile spherical neutron detector. These measurements are done in the reactor hall around the horizontal experimental channels. Measured values of neutron radiation are three times lower than the relevant levels of gamma radiation [sr

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Isothermal calorimeter for reactor radiation dosimetry

Energy Technology Data Exchange (ETDEWEB)

Radak, B; Markovic, V [Institute of Nuclear Sciences Boris Kidric, Odeljenje za radijacionu hemiju, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

An isothermal calorimeter with thermistors for measuring absorbed dose rates from 10{sup 4}-5-6.10{sup 5} rad/h in reactor experimental holes has been designed. A kinetics method for determining the equilibrium temperature difference has been developed, and its application in isothermal calorimetry proved. The expected accuracy in measurements within {+-} 2-5% has been proved by measurements carried out in the reactor. Some data obtained by measurements in the reactor RA are presented (author)

Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

International Nuclear Information System (INIS)

Lai, W.; Collins, E.K.

1977-01-01

This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

What occurred in the reactors

International Nuclear Information System (INIS)

Kudo, Kazuhiko

2013-01-01

Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

International Nuclear Information System (INIS)

Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

1989-01-01

Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

International Nuclear Information System (INIS)

Brand, P.; Wiechers, W.K.

1974-01-01

The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

A study of operation of station fire brigade with comparison for Japan and the U.S

International Nuclear Information System (INIS)

Ichiki, Kuniyasu

2011-01-01

With the occurrence of a fire at Kashiwazaki Kariwa No. 3 unit on a site transformer when Niigata Chuetsu Offshore Earthquake occurred in 2007, the movement of establishing station fire brigade broke out all over Japan. Fire brigade of U.S. power stations is an organization structured with the standards of clear requirements. Some of station fire event reports include examples of activation of station fire brigade. Preparing station fire brigade with good expertise is quite meaningful on the viewpoint of ruling out significant threats of fire against nuclear safety, such as Browns Ferry station fire occurred in 1975. The purpose of this study is to extract useful knowledge through making research on examples of operations of station fire brigade of both Japan and the U.S. As for fire occurrence, relatively frequent fires are seen at Japanese stations with miscellaneous components or some material for works. In the U.S., on the other hand, we can identify relatively frequent fires occurred with important equipment. This might make fire brigades of the U.S. well-prepared for fire fighting by establishing fire-coping strategies for every location in the station. Also, in some stations, fire brigade is continuously trained to enhance its firefighting and rescuing abilities. With the aspect of training of brigades, they are often assembled with devoted staff, which enable them acquiring frequent training and educations to improve their skills further. Fire fighting and rescuing are inevitably dangerous task. This is why frequent training is essential in order to prevent injuries and to keep up their motivation. Unfortunately, the circumstance for brigades in Japan is not helpful enough in experiencing special training needed for them. We have to figure out some way to get over such disadvantages. (author)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor, United States of America

Energy Technology Data Exchange (ETDEWEB)

Boing, L E; Bhattacharyya, S K [Technology Development Division, Decommissioning Program, Argonne National Laboratory, Argonne, IL (United States)

2002-02-01

The aging of research reactors worldwide has resulted in a heightened awareness in the international decommissioning community of the timeliness to review and address the needs of research reactor operators in planning for and eventually performing the decommissioning of these types of facilities. Many reactors already undergoing decommissioning can be used as test beds for evaluating enhanced or new/innovative technologies for decommissioning; it is possible that new techniques could be made available for future research reactor-decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the decommissioners in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to their use in actual research reactor decommissioning. The decommissioning of the CP-5 Research Reactor located at the ANL-East Site has been completed. In this paper we present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors. In addition, details are provided on other related U.S. D and D activities, which may be useful to the international research reactor D and D community. (author)

Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor, United States of America

International Nuclear Information System (INIS)

Boing, L.E.; Bhattacharyya, S.K.

2002-01-01

The aging of research reactors worldwide has resulted in a heightened awareness in the international decommissioning community of the timeliness to review and address the needs of research reactor operators in planning for and eventually performing the decommissioning of these types of facilities. Many reactors already undergoing decommissioning can be used as test beds for evaluating enhanced or new/innovative technologies for decommissioning; it is possible that new techniques could be made available for future research reactor-decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the decommissioners in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to their use in actual research reactor decommissioning. The decommissioning of the CP-5 Research Reactor located at the ANL-East Site has been completed. In this paper we present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors. In addition, details are provided on other related U.S. D and D activities, which may be useful to the international research reactor D and D community. (author)

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

International Nuclear Information System (INIS)

Brooks, B.G.

1985-03-01

This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

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.

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

A multi-purpose reactor

International Nuclear Information System (INIS)

Changwen Ma

2000-01-01

An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

Energy Technology Data Exchange (ETDEWEB)

Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

2006-02-15

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 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang 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 Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.

1989-01-01

The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

Farewell to a reactor

International Nuclear Information System (INIS)

Skanborg, P.

1976-01-01

Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Energy Technology Data Exchange (ETDEWEB)

El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

2017-11-15

In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

Brazilian multipurpose reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

Brazilian multipurpose reactor

International Nuclear Information System (INIS)

2014-01-01

The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U 3 Si 2 dispersion-type Al having a density of up to 3.5 gU/cm 3 and enrichment of 19.75% by weight of 23 5 U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive facilities are

Analyzing the loss of coolant accident in PWR nuclear reactors with elevation change in cold leg by RELAP5/MOD3.2 system code

International Nuclear Information System (INIS)

Kheshtpaz, H.; Alison, C.

2006-01-01

As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the RELAP5/MOD3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect

Technical mechanics in constructional reactor safety

International Nuclear Information System (INIS)

Matthees, W.

1979-01-01

Reactor safety is based on close cooperation between a number of technical and scientific disciplines; most problems of reactor technology can be solved with the aid of technical mechanics. At the 5th International Conference on Structural Mechanics in Reactor Technology (5th SMIRT), one of the biggest conferences in the field of applied technical mechanics, about 800 papers were read giving the latest state of knowledge in the field of constructional reactor safety. The main subject of the conference was the analysis of material behaviour under high loads; the information and methods of these analysis go far beyond what is required in the conventional field. (orig./UA) [de

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D; Estudo termofluidodinâmico de reatores nucleares avançados de alta temperatura utilizando o RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Scari, Maria Elizabeth

2017-07-01

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{sub 2}, the LiF-BeF{sub 2}, also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO

Task 24: Dynamic analysis of Kozloduy NPP unit 5 structures: Reactor building

International Nuclear Information System (INIS)

Zola, M.

1999-01-01

This report refers to the activities of a sub-contract to the Project RER/9/046, awarded to ISMES by the International Atomic Energy Agency (IAEA) of Vienna, to compare the results obtained from the experimental activities performed under previous contract by ISMES with those coming from analytical studies performed in the framework of the Coordinated Research Programme (CRP) on 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants' by other Institutions, relevant to Kozloduy Unit 5 reactor building. After a brief introduction to the problem in Chapter 1, the identification of the comparison positions and reference directions is given in Chapter 3. A very quick description of the performed experimental tests is given in Chapter 4, whereas the characteristics of both experimental and analytical data are presented in Chapter 5. The data processing procedures are reported in Chapter 6 and some simple remarks are given in Chapter 7. (author)

Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2012-01-01

Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

RA Reactor applications, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Cupac, S.; Stanic, A.

1990-01-01

RA reactor was not operated during the past five years due to the renewal and reconstruction of the reactor systems, which in underway. In the period from 1986-1990, reactor was operated only 144 MWh in 1986, for the need of testing the reactor systems and possibility of irradiating 125 I. Reactor will not be operated in 1991 because of the exchange of complete instrumentation which is planned to be finished by the end of 1991. It is expected to start operation in May 1992. That is why this annex includes the plan of reactor operation for period of nine months starting from from the moment of start-up. It is planned to operate the reactor at 0.02 MW power first three months, to increase the power gradually and reach 3.5 MW after 8 months of operation. It is foreseen to operate the reactor at 4.7 MW from the tenth month on [sr

Engineering reactors for catalytic reactions

Indian Academy of Sciences (India)

126, No. 2, March 2014, pp. 341–351. c Indian Academy of Sciences. ... enhancement was realized by catalyst design, appropriate choice of reactor, better injection and .... Gas–liquid and liquid–solid transport processes in catalytic reactors.5.

Future view of total energy system and reactor engineering and reactor physics

International Nuclear Information System (INIS)

Ozawa, T.

1974-01-01

This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

Production of Advanced Biofuels via Liquefaction - Hydrothermal Liquefaction Reactor Design: April 5, 2013

Energy Technology Data Exchange (ETDEWEB)

Knorr, D.; Lukas, J.; Schoen, P.

2013-11-01

This report provides detailed reactor designs and capital costs, and operating cost estimates for the hydrothermal liquefaction reactor system, used for biomass-to-biofuels conversion, under development at Pacific Northwest National Laboratory. Five cases were developed and the costs associated with all cases ranged from $22 MM/year - $47 MM/year.

Main safety lessons from 5-year operation of the renovated Dalat nuclear research reactor

International Nuclear Information System (INIS)

Anh, T.H.; Lam, P.V.; An, T.K.; Khang, N.P.; Tan, D.Q.

1989-01-01

The paper presents main safety related characteristics of the Dalat Nuclear Research Reactor (DNRR), which was reconstructed in 1982 at the site of the former TRIGA Mark II, while retaining some of its structures. Experience acquired from reactor operation is analysed. The programme of investigations aimed at better ensuring nuclear safety of the reactor, together with some of its results are presented. Finally some propositions to improve the present situation are suggested. (Authors). (2 Tables, 2 fig.)

Reactor Sharing Program

International Nuclear Information System (INIS)

Tehan, Terry

2002-01-01

Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

International Nuclear Information System (INIS)

Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

1974-01-01

In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

FBR type reactor

International Nuclear Information System (INIS)

Yamaoka, Mitsuaki

1988-01-01

Purpose: To enable to increase the burning period by enabling to decrease the reduction of burning reactivity and unifying the irradiation amount of fast neutrons. Constitution: A cylindrical reactor core made of fissile material-enriched fuel is constituted so as to form a plurality of layer-like enriched regions in which the enrichment degree of the fissile material is increased from the center to the radial and axial directions. Then, the ratio between the average enrichment degree for all of the enrichment regions other than the region at the reactor core center with the lowest enrichment degree and the enrichment degree of the enriched region formed at the center of the reactor core is made greater by 5 % or 20 % than the ratio at the initial burning stage where the power distribution of the reactor core is most flattened. (Kawakami, Y.)

Report on the specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors organized by the international atomic energy agency at Oarai Engineering Centre of power reactor and nuclear development corporation, Japan, November 5-7, 1991

International Nuclear Information System (INIS)

Paranjpe, S.R.

1992-01-01

As recommended by the International Working Group on Fast Reactors (IWGFR), the International Atomic Energy Agency organized a specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors. Specialists from all member countries of IWGFR-China, France, Germany, India, Italy, Japan, Russia, the United Kingdom, and the United States-participated in the meeting and made presentations as listed in Table 1. The Commission of European Communities also sent representatives to the meeting. Table 2 contains a list of participants. The meeting consisted of five sessions: (1) an overview, (2) safety characteristics of decay heal removal systems, (3) safely characteristics of reactor protection systems and reactor shutdown systems, (4) safely characteristics of reactor cores, and (5) general discussions antiformulation of recommendations

Predicting the in-reactor mechanical behavior of Zr-2.5Nb pressure tubes from postirradiation microstructural examination data

International Nuclear Information System (INIS)

Griffiths, M.; Davies, P.H.; Davies, W.G.; Sagat, S.

2002-01-01

Postirradiation microstructure examinations of Zr-2.5Nb pressure tubes removed from service in CANDU reactors have shown clear trends in the dislocation structure and the state of the β-phase, as a function of operating temperature, neutron flux, and time. These microstructural parameters correlate well with changes in the mechanical properties. For example, the rapid increase in dislocation loop density in the early stages of irradiation corresponds with a rapid increase in tensile strength and DHC velocity, and a corresponding decrease in fracture toughness. There is also a strong negative correlation between the degree of β-phase decomposition and DHC velocity. In addition to the effects of microstructure evolution on the mechanical properties, changes in the a-type and c-component dislocation loop densities also affect irradiation deformation (creep and growth). Statistical analyses of the irradiation microstructure data have been used to derive empirical relationships between dislocation densities and β-phase structure with temperature, flux, and time. The relationships thus derived are useful in predicting where the mechanical properties are most affected by the in-reactor operating conditions. The predictions are compared with mechanical test data for samples from various axial and circumferential locations of 42 pressure tubes removed from operating CANDU reactors. The results are discussed in terms of the mechanisms controlling tensile strength, fracture, delayed-hydride-cracking, and in-reactor deformation. (author)

Dosimetry and radiation shielding at the RA reactor, Annual report 1975, Annex 5; Prilog 5 - Dozimetrija i tehnicka zastita

Energy Technology Data Exchange (ETDEWEB)

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

1976-01-15

In the working environment at the RA reactor, the level of gamma radiation is measured continuously by the built-in stationary system. According to the needs, measurement are done in the reactor hall every day. The level of gamma radiation is measured separately in typical points when the reactor is operated at nominal power and during intervals between two operating campaigns. The level of neutron radiation is measured according to the needs by means of a mobile spherical neutron detector. These measurements are done in the reactor hall around the horizontal experimental channels. Measured values of neutron radiation are three times lower than the relevant levels of gamma radiation. [Serbo-Croat] Na reaktoru RA vrsi se stalna kontrola nivoa gama zracenja po tehnoloskim i radnim prostorijama, pomocu ugradjenog stacionarnog sistema. Svakodnevno se vrse merenja u reaktorskoj hali, prema ukazanim potrebama. Posebno se mere nivoi gamma zracenja u karakteristicnim tackama pri radu reaktora na nominalnoj snazi i u pauzama izmedju dva kampanje. Merenje nivoa neutronskog zracenja vrsi se diskontinualno pomocu mobilnog sfernog neutronskog dozimetra. Merenja se obavljaju u hali oko horizontalnih eksperimentalnih kanala. Izmerene vrednosti su u proseku tri puta manje od odgovarajucih nivoa gama zracenja.

Analysis of reactivity transient for the DIDO type research reactors using RELAP5

International Nuclear Information System (INIS)

Adorni, M.; Bousbia-Salah, A.; D'Auria, F.; Nabbi, R.

2005-01-01

Recent availability of high performance computers and computational methods together with the continuing increase in operational experience imposes revising some operational constrains and conservative safety margins. The application of Best-Estimate (BE) method constitutes a real necessity in the safety and design analysis and allows getting more realistic simulation of the processes taking place during the steady state operation and transients. In comparison to the conservative approaches, the application of Best-Estimate methods results in the mitigation of the constraining limits in design and operation. This paper presents the results of the application of the RELAP5/Mod3.3 system thermal-hydraulic code to the German FRJ-2 research reactor for a reactivity transient, which has been analyzed in the past using the verified system code CATHENA [1], [2], [3]. The work mainly aims checking the capability of RELAP5 [4] for research reactor transient analysis by the comparison of the results of the two codes and including modeling basis and analytical approaches. According to the existing references RELAP5 applications are concentrated on the transient analysis of nuclear power systems. The considered case consists of a simulation related to a hypothetical fast reactivity transient, which is assumed to be caused by the failure of one shutdown arm. The case has been chosen due to the importance of the models for the precise description of the complex phenomenon of subcooled boiling and two phase flow taking place during the transient. For this purpose, the fuel element assembly was modeled in detail according to design data. The primary circuit was included in the whole model in order to consider the interaction with individual fuel elements with core. In general the results of the two codes are in agreement and comparable during the initial phase of the transient. After reaching the flow regime with fully developed nucleate boiling and two phase flow RELAP5 exhibits

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

Energy Technology Data Exchange (ETDEWEB)

Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

2015-10-15

In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

Application of RELAP5-3D code for thermal analysis of the ADS reactor core; Aplicação do código RELAP5-3D para análise térmica do núcleo de um reator ADS

Energy Technology Data Exchange (ETDEWEB)

Fernandes, Gustavo Henrique Nazareno

2018-04-01

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

Turning points in reactor design

International Nuclear Information System (INIS)

Beckjord, E.S.

1995-01-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems

Turning points in reactor design

Energy Technology Data Exchange (ETDEWEB)

Beckjord, E.S.

1995-09-01

This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5

International Nuclear Information System (INIS)

Pena V, J.D.

2016-01-01

The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

Thorium utilisation in thermal reactors

International Nuclear Information System (INIS)

Balakrishnan, K.

1997-01-01

It is now more or less accepted that the best way to use thorium is in thermal reactors. This is due to the fact that U233 is a good material in the thermal spectrum. Studies of different thorium cycles in various reactor concepts had been carried out in the early days of nuclear power. After three decades of neglect, the world is once again looking at thorium with some interest. We in India have been studying thorium cycles in most of the existing thermal reactor concepts, with greater emphasis on heavy water reactors. In this paper, we report some of the work done in India on different thorium cycles in the Indian pressurized heavy water reactor (PHWR), and also give a description of the design of the advanced heavy water reactor (AHWR). (author). 1 ref., 2 tabs., 5 figs

DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

Directory of Open Access Journals (Sweden)

Ralind Re Marla

2015-03-01

Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

Iris reactor conceptual design

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Reactor utilization; Eksploatacija reaktora

Energy Technology Data Exchange (ETDEWEB)

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

1963-02-15

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

Reactor for exothermic reactions

Science.gov (United States)

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Comparison of SERPENT and CASMO-5M for pressurized water reactors models

International Nuclear Information System (INIS)

Hursin, M.; Vasiliev, A.; Ferroukhi, H.; Pautz, A.

2013-01-01

The objective of this work is to perform a preliminary assessment of the capability of SERPENT to generate cross sections for a PWR Beginning-of-Life (BOL) isothermal mini-core by comparing a SERPENT/PARCS solution with the results obtained using a CASMO-5M/PARCS approach. The PARCS code is used instead of the usual SIMULATE-3 to analyze the Swiss Reactors, because interfaces with PARCS already exist to obtain neutronic data from SERPENT. For the PWR configurations, the differences between CASMO-5M and SERPENT solutions are within 200 pcm at the assembly level and thus rather small when considering the deterministic transport method (energy/angular/space discretization) in CASMO-5M versus the stochastic treatment of SERPENT, the statistical uncertainties in the Monte-Carlo approach as well as the eventual differences in nuclear data used by both codes. At the 2D mini-core level, no major difference is observed when comparing PARCS run with CASMO-5M versus SERPENT cross sections. For the generation of kinetic parameters, non trivial differences are observed due both to the methods and the data used. For the relatively limited number of configurations considered, it is hard to make any definitive statement on the benefits of using Monte Carlo codes in terms of nuclear data generation. (authors)

Cooperation of nuclear reactor controller ARM-5S and turbine TVER-02

International Nuclear Information System (INIS)

Wagner, K.; Lnenicka, B.; Pokorny, F.; Prochazka, F.

1985-01-01

Turbines of Czechoslovak make provided with controllers TVER-02 are installed in WWER-440 nuclear power plants under construction in Czechoslovakia. Reactor output is controlled using Soviet-made controllers ARM-5S which already comprise turbine controllers. The problems are analyzed of cooperation of both controllers, especially their parameters and transient processes in typical operating situations. The analysis uses the results of measurements performed during the power start-up of Unit 1 of the V-2 nuclear power plant at Jaslovske Bohunice. The results show that two types of control modes can be selected for the operation of the entire unit: the control to constant unit output, and control of unit output varying with turbine load selected on the TVER-02 controller or given by the demand of the power network. (Z.M.)

First TREAT [Transient Reactor Test Facility] transient overpower tests on U-Pu-Zr fuel: M5 and M6

International Nuclear Information System (INIS)

Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

1987-01-01

Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported [1,2]. Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

International Nuclear Information System (INIS)

Ojeda S, J.; Morales S, J.; Chavez M, C.

2009-10-01

This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

International Nuclear Information System (INIS)

Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

1977-01-01

An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

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

1989-01-01

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

The Integral Fast Reactor

International Nuclear Information System (INIS)

Till, C.E.; Chang, Y.I.; Lineberry, M.J.

1990-01-01

Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs

Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

International Nuclear Information System (INIS)

Hanan, N.A.; Liaw, J.R.; Matos, J.E.

2005-01-01

A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.; Miller, J.R.

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

The contract of design of an atomic reactor system in Yonggwang - 5, 6 nuclear power plant

International Nuclear Information System (INIS)

1995-05-01

This is a contract of design of an atomic reactor system in Yonggwang 5, 6 nuclear power plant. It has the general contract condition. In the appendix, it indicates the detail regulations between two parties which are the coverage of division on the responsibility, schedule of the delivery, standard of the technology, guarantee, drawing and paper support of the Korea Electric Power Corporation, support of technology drill, test, regulations of code and standard and list of items and prices.

Submersion-Subcritical Safe Space (S4) reactor

International Nuclear Information System (INIS)

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

2006-01-01

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

Fusion reactor safety

International Nuclear Information System (INIS)

1987-12-01

Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

Energy Technology Data Exchange (ETDEWEB)

Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, P.O. BOX 1415, Zanjan (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, P.O. BOX 212 Tehran (Iran, Islamic Republic of); Kiai, S.M. Sadat [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I. P.O. BOX 14155-1339, Tehran (Iran, Islamic Republic of)], E-mail: sadatkiai@yahoo.com

2009-02-15

A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity ({rho}), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.

Physical security at research reactors

International Nuclear Information System (INIS)

Clark, R.A.

1977-01-01

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

Planning the Decommissioning of Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

2013-08-15

In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)

TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

International Nuclear Information System (INIS)

Solbrig, C.W.; Stevens, W.W.

1990-01-01

Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Operational safety and reactor life improvements of Kyoto University Reactor

International Nuclear Information System (INIS)

Utsuro, M.; Fujita, Y.; Nishihara, H.

1990-01-01

Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

Simulation development for TRIGA reactor

International Nuclear Information System (INIS)

Handoyo, D.

1997-01-01

A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

Nuclear reactor multi-physics simulations with coupled MCNP5 and STAR-CCM+

International Nuclear Information System (INIS)

Cardoni, Jeffrey Neil; Rizwan-uddin

2011-01-01

The MCNP5 Monte Carlo particle transport code has been coupled to the computational fluid dynamics code, STAR-CCM+, to provide a high fidelity multi-physics simulation tool for pressurized water nuclear reactors. The codes are executed separately and coupled externally through a Perl script. The Perl script automates the exchange of temperature, density, and volumetric heating information between the codes using ASCII text data files. Fortran90 and Java utility programs assist job automation with data post-processing and file management. The MCNP5 utility code, MAKXSF, pre-generates temperature dependent cross section libraries for the thermal feedback calculations. The MCNP5–STAR-CCM+ coupled simulation tool, dubbed MULTINUKE, was applied to a steady state, PWR cell model to demonstrate its usage and capabilities. The demonstration calculation showed reasonable results that agree with PWR values typically reported in literature. Temperature and fission reaction rate distributions were realistic and intuitive. Reactivity coefficients were also deemed reasonable in comparison to historically reported data. The demonstration problem consisted of 9,984 CFD cells and 7,489 neutronic cells. MCNP5 tallied fission energy deposition over 3,328 UO_2 cells. The coupled solution converged within eight hours and in three MULTINUKE iterations. The simulation was carried out on a 64 bit, quad core, Intel 2.8 GHz microprocessor with 1 GB RAM. The simulations on a quad core machine indicated that a massively parallelized implementation of MULTINUKE can be used to assess larger multi-million cell models. (author)

RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

Energy Technology Data Exchange (ETDEWEB)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

1995-09-01

This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

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

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.)

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)

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

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

International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Blevins, J.D.; Stasko, R.R.

1989-09-01

An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

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

Prospects for applications of ship-propulsion nuclear reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.

1994-01-01

The use of ship-propulsion nuclear power reactors in remote areas of Russia is examined. Two ship reactors were analyzed: the KLT-40, a 170 MW-thermal reactor; and the KN-3, a 300 MW-thermal reactor. The applications considered were electricity generation, desalination, and drinking water production. Analyses showed that the applications are technically justified and could be economically advantageous. 5 refs., 9 figs., 1 tab

Maintenance and material aspects of DREAM reactor

International Nuclear Information System (INIS)

Ueda, S.; Nishio, S.; Yamada, R.; Seki, Y.; Kurihara, R.; Adachi, J.; Yamazaki, S.

2000-01-01

A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor

Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

Energy Technology Data Exchange (ETDEWEB)

Lai, W.; McCauley, E.W.

1978-01-04

Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90/sup 0/ torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this.

Air scaling and modeling studies for the 1/5-scale mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Lai, W.; McCauley, E.W.

1978-01-01

Results of table-top model experiments performed to investigate pool dynamics effects due to a postulated loss-of-coolant accident (LOCA) for the Peach Bottom Mark I boiling water reactor containment system guided subsequent conduct of the 1/5-scale torus experiment and provided new insight into the vertical load function (VLF). Pool dynamics results were qualitatively correct. Experiments with a 1/64-scale fully modeled drywell and torus showed that a 90 0 torus sector was adequate to reveal three-dimensional effects; the 1/5-scale torus experiment confirmed this

Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

International Nuclear Information System (INIS)

Feng, K.M.; Huang, J.H.

1994-01-01

The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

Thermionic reactors for space nuclear power

Science.gov (United States)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Fusion reactors as a future energy source

International Nuclear Information System (INIS)

Seifritz, W.

A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

Energy Technology Data Exchange (ETDEWEB)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

1996-02-01

In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

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.)

Vibration-proof FBR type reactor

International Nuclear Information System (INIS)

Kawamura, Yutaka.

1992-01-01

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

Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

International Nuclear Information System (INIS)

Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

1981-01-01

Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

Overview of fusion reactor safety

International Nuclear Information System (INIS)

Cohen, S.; Crocker, J.G.

1981-01-01

Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

Research reactor spent fuel in Ukraine

International Nuclear Information System (INIS)

Trofimenko, A.P.

1996-01-01

This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

RELAP5/MOD2. 5 analysis of the HFBR (High Flux Beam Reactor) for a loss of power and coolant accident

Energy Technology Data Exchange (ETDEWEB)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs.

RELAP5/MOD2.5 analysis of the HFBR [High Flux Beam Reactor] for a loss of power and coolant accident

International Nuclear Information System (INIS)

Slovik, G.C.; Rohatgi, U.S.; Jo, Jae.

1990-05-01

A set of postulated accidents were evaluated for the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory. A loss of power accident (LOPA) and a loss of coolant accident (LOCA) were analyzed. This work was performed in response to a DOE review that wanted to update the understanding of the thermal hydraulic behavior of the HFBR during these transients. These calculations were used to determine the margins to fuel damage at the 60 MW power level. The LOPA assumes all the backup power systems fail (although this event is highly unlikely). The reactor scrams, the depressurization valve opens, and the pumps coast down. The HFBR has down flow through the core during normal operation. To avoid fuel damage, the core normally goes through an extended period of forced down flow after a scram before natural circulation is allowed. During a LOPA, the core will go into flow reversal once the buoyancy forces are larger than the friction forces produced during the pump coast down. The flow will stagnate, reverse direction, and establish a buoyancy driven (natural circulation) flow around the core. Fuel damage would probably occur if the critical heat flux (CHF) limit is reached during the flow reversal event. The RELAP5/MOD2.5 code, with an option for heavy water, was used to model the HFBR and perform the LOPA calculation. The code was used to predict the time when the buoyancy forces overcome the friction forces and produce upward directed flow in the core. The Monde CHF correlation and experimental data taken for the HFBR during the design verification phase in 1963 were used to determine the fuel damage margin. 20 refs., 40 figs., 11 tabs

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)

Proceedings of the advisory committee on reactor safeguards workshop on future reactors

International Nuclear Information System (INIS)

2001-12-01

This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda

Proceedings of the advisory committee on reactor safeguards workshop on future reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-12-01

This report contains the information presented at the Advisory Committee on Reactor Safeguards Workshop on Future Reactors held at the Nuclear Regulatory Commission headquarters in Rockville, Maryland, on June 4-5, 2001. Included are the subject matter summaries, followed by the presentation material and selected participants discussions. The primary purpose of the workshop was to identify the regulatory challenges associated with future reactor designs. A list of such challenges was developed from the workshop notes, the various presentations, the panel discussions and the question and answer sessions. This list is included in the Introduction section of this document. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final workshop agenda.

SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

International Nuclear Information System (INIS)

Harvego, E. A.; Siefken, L. J.

2000-01-01

The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Backfitting of the FRG reactors

Energy Technology Data Exchange (ETDEWEB)

Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

1990-05-01

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

Backfitting of the FRG reactors

International Nuclear Information System (INIS)

Krull, W.

1990-01-01

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

Generation IV reactors: international projects

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

2003-01-01

Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

Practical course on reactor instrumentation

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-06-01

This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

Compact toroid refueling of reactors

International Nuclear Information System (INIS)

Gouge, M.J.; Hogan, J.T.; Milora, S.L.; Thomas, C.E.

1988-04-01

The feasibility of refueling fusion reactors and devices such as the International Thermonuclear Engineering Reactor (ITER) with high-velocity compact toroids is investigated. For reactors with reasonable limits on recirculating power, it is concluded that the concept is not economically feasible. For typical ITER designs, the compact toroid fueling requires about 15 MW of electrical power, with about 5 MW of thermal power deposited in the plasma. At these power levels, ideal ignition (Q = ∞) is not possible, even for short-pulse burns. The pulsed power requirements for this technology are substantial. 6 ref., 1 figs

Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

2000-01-01

U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

RA Research reactor, Annual report 1971

International Nuclear Information System (INIS)

Milosevic, D. et al.

1971-12-01

During 1971, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and 50 days at lower power levels. Total production mounted to 31606 MWh which is 5.3% higher than planned, and the highest annual level since the reactor started operation. Reactor was used for irradiation and experiments according to the demand of 425 users, of which 370 from the Institute and 55 external users. This report contains detailed data about reactor power and experiments performed in 1971. Discrepancies from the action plan, meaning higher production was achieved in June and December due to special demand of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to power cuts during reactor operation. There was no longer interruption caused by failures of the equipment. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. Shorter interruptions resulted from breaking of connectors in the technical water pipe system caused by soil sliding near the pumping station on the Danube. Total personnel exposure dose was lower than during previous years. There was no accident nor any event that could be called accidental. Decontamination od surfaces was less than during previous years. It was concluded that the successful operation in 1971 resulted from efficient work during past years. But, some of the activities were interrupted due to undefined policy concerned with operation of the RA reactor and financial issues. This involves study of the possibility to use highly enriched fuel that would increase the useful neutron flux and the reactor compatibility with similar reactors for the future ten years. Another project that has been interrupted is construction of the emergency core cooling system which is important for the reactor safety. Financial problems have influenced not only the reactor operation but the number of employees, which could cause negative

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

Directory of Open Access Journals (Sweden)

Baschwitz Anne

2016-01-01

Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

International Nuclear Information System (INIS)

1968-01-01

Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

International Nuclear Information System (INIS)

Hagrman, D.T.

1995-06-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident

Water treatment process for nuclear reactors

International Nuclear Information System (INIS)

Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

1992-01-01

Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerates to obtain the optimum quantity of pure water which reached to 15 cubic meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30 %. output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined.5 fig., 3 tab

Permeated defect detecting test method and device in reactor

International Nuclear Information System (INIS)

Sakurai, Yoshishige.

1996-01-01

The present invention provides a method of and a device capable of performing a test for entire inner surfaces of the reactor upon periodical inspection of a BWR type reactor while sufficiently taking countermeasures for radiation rays into consideration. Namely, the present invention comprises following steps. (1) A provisional step for taking a shroud head of a reactor core shroud and incore structural components above and below the shroud out of the reactor, discharging reactor water and water tightly closing openings such as reactor wall perforation holes, (2) a pretreatment step for washing exposed inner surfaces of the reactor and peeling deteriorated materials, (3) a first drying step for drying portions washed and peeled in the step (2), (4) a permeation step for applying a permeation liquid of a defect detecting medium on the exposed inner surfaces of the reactor, (5) a permeation liquid removing step for removing the an excess permeation liquid in the step (4), (6) a second drying step for drying corresponding portions after performing the step (5), and (7) a flaw detecting step for optically observing the corresponding portions after performing the step (6) and detecting flaws. (I.S.)

MLR reactor

International Nuclear Information System (INIS)

Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

1998-01-01

The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

International Nuclear Information System (INIS)

Freels, J.D.

1993-01-01

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed

Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

International Nuclear Information System (INIS)

Rubini, L.A.

1978-02-01

An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

International Nuclear Information System (INIS)

Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

2012-01-01

The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

Backfitting swimming pool reactors

International Nuclear Information System (INIS)

Roebert, G.A.

1978-01-01

Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2000-01-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2000-07-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

Energy Technology Data Exchange (ETDEWEB)

Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

2016-03-15

In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

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

2003-01-01

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

Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

International Nuclear Information System (INIS)

Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying; Lee, Hyung-Sool; Wang, Ai-Jie

2012-01-01

Highlights: ► A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. ► Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). ► PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. ► The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 ± 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m −3 d −1 ) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m −3 d −1 (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

Energy Technology Data Exchange (ETDEWEB)

Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

2012-11-15

Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

Nuclear reactors to come

International Nuclear Information System (INIS)

Lung, M.

2002-01-01

The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

A friendly Maple module for one and two group reactor model

International Nuclear Information System (INIS)

Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O.

2015-01-01

The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

A friendly Maple module for one and two group reactor model

Energy Technology Data Exchange (ETDEWEB)

Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O., E-mail: camila.oliv.baptista@gmail.com, E-mail: pavanguilherme@gmail.com, E-mail: kelmo.lins@gmail.com, E-mail: marcelovilelasilva@gmail.com, E-mail: rodrigowerner@hotmail.com, E-mail: neutron201566@yahoo.com, E-mail: vellozo@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

2015-07-01

The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

TREAT Reactor Control and Protection System

International Nuclear Information System (INIS)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

Assessment of torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Painter, S.L.

1992-12-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R 0 = 6.6-8.8 m, on-axis magnetic field B 0 = 4.8-7.5 T, B max (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions

Consequence analysis for nuclear reactors, Yongbyon

International Nuclear Information System (INIS)

Kang, Taewook; Jae, Moosung

2017-01-01

Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea. (author)

Designing a mini subcritical nuclear reactor

International Nuclear Information System (INIS)

Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M.

2015-10-01

In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of 239 PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of 239 PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k e -f f , the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k eff , the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons 239 PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k eff was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k eff of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five cylinders not met. (Author)

Remote dismantlement tasks for the CP5 reactor: Implementation, operations, and lessons learned

International Nuclear Information System (INIS)

Noakes, M.W.

1998-01-01

This paper presents a developer's perspective on lessons learned from one example of the integration of new prototype technology into a traditional operations environment. The dual arm work module was developed by the Robotics Technology Development Program as a research and development activity to examine manipulator controller modes and deployment options. It was later reconfigured for the dismantlement of the Argonne National Laboratory Chicago Pile number-sign 5 reactor vessel as the crane-deployed dual arm work platform. Development staff worked along side operations staff during a significant part of the deployment to provide training, maintenance, and tooling support. Operations staff completed all actual remote dismantlement tasks. At the end of available development support funding, the Dual Arm Work Platform was turned over to the operations staff, who is still using it to complete their dismantlement tasks

Nuclear reactor

International Nuclear Information System (INIS)

Hattori, Sadao; Sato, Morihiko.

1994-01-01

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

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Savannah River Site production reactor safety analysis report

International Nuclear Information System (INIS)

1996-01-01

The process water system (PWS) is designed to remove heat produced in the reactor from the fission process, gamma radiation absorption, and fission product decay. Heat removal is accomplished by circulating heavy water through the reactor. Cooling is provided for fuel assemblies, target assemblies, control rods, bulk moderator, deflector plate, reactor tank, and reactor structural components. Approximately 90% of the heat load is generated in the fuel and target assemblies, 5% in the moderator, and 5% in the shielding. In addition to serving as the-heat transfer medium, the process water moderates neutrons produced by fission in the fuel. D 2 O is used in this application because of its favorable moderating and neutron capture properties, which result in high neutron efficiency and reactor productivity. The PWS piping and components also provide a high-integrity leak barrier against loss of moderator and the radioactive fission and corrosion products. Components of the PWS are located in the reactor building between the -40-foot elevation and the 0-foot elevation. Specific locations include the process room, heat exchanger bay, motor rooms, and pump rooms. The system diagram is shown in Figure 5.1-2. PWS design data are presented in Table 5.1-1. The PWS consists of six parallel heat transfer loops. In each loop, approximately 25,000 gpm of D 2 O is circulated from one of six outlet nozzles in the bottom of the reactor tank through a motor-operated valve (MOV) to the suction side of the process water pump. Each pump is driven by an AC motor and a DC motor through a gear reducer unit. A 3-ton flywheel on the drive shaft of the AC motor provides gradual flow coastdown when power is lost. During reactor operation, the DC motors are operated continuously from the diesel generator sets as backup to the AC motors. Following shutdown, the DC motors are operated to provide adequate circulation and core cooling

WWER-440 type reactor core

International Nuclear Information System (INIS)

Mizov, J.; Svec, P.; Rajci, T.

1987-01-01

Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

International Nuclear Information System (INIS)

Obara, Toru; Sekimoto, Hiroshi

1997-01-01

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

Chemical surveillance of commercial fast breeder reactors

International Nuclear Information System (INIS)

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

1988-01-01

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

Seventh regular meeting of the International Working Group on Reliability of Reactor Pressure Components, Vienna, 3-5 September 1985

International Nuclear Information System (INIS)

1986-07-01

The seventh regular meeting of the IAEA International Working Group on Reliability of Reactor Pressure Components was held at the Agency's Headquarters in Vienna from 3 to 5 September 1985. The representatives of Member States and of the Commission of the European Communities reported the status of the research programmes in this field (12 presentations). A separate abstract was prepared for each of the presentations

Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics; Analiza sigurnosti rada reaktora RA - I-III, I deo - Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 10{sup 13} neutrons/cm{sup 2} s, and maximum neutron flux 5.5 10{sup 13} neutrons/cm{sup 2} s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume.

Savannah River Site reactor hardware design modification study

International Nuclear Information System (INIS)

Fisher, J.E.

1990-01-01

A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR): Data manual. Part 3: Hardware component failure data; Volume 5, Revision 4

International Nuclear Information System (INIS)

Reece, W.J.; Gilbert, B.G.; Richards, R.E.

1994-09-01

This data manual contains a hard copy of the information in the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) Version 3.5 database, which is sponsored by the US Nuclear Regulatory Commission. NUCLARR was designed as a tool for risk analysis. Many of the nuclear reactors in the US and several outside the US are represented in the NUCLARR database. NUCLARR includes both human error probability estimates for workers at the plants and hardware failure data for nuclear reactor equipment. Aggregations of these data yield valuable reliability estimates for probabilistic risk assessments and human reliability analyses. The data manual is organized to permit manual searches of the information if the computerized version is not available. Originally, the manual was published in three parts. In this revision the introductory material located in the original Part 1 has been incorporated into the text of Parts 2 and 3. The user can now find introductory material either in the original Part 1, or in Parts 2 and 3 as revised. Part 2 contains the human error probability data, and Part 3, the hardware component reliability data

Reactor types for the future

International Nuclear Information System (INIS)

Hall, A.C.

1990-01-01

The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

Reactor types for the future

Energy Technology Data Exchange (ETDEWEB)

Hall, A C [PWR Power Projects Ltd., Knutsford, Cheshire (United Kingdom)

1990-06-01

The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

Fuel performance update at Japanese BWRs

International Nuclear Information System (INIS)

Otsuka, Yasuyuki; Abe, Moriyasu

2008-01-01

Fuel Performance of Japanese BWRs has shown excellency with a very low fuel failure rate over the past couple of decades. In order to eliminate debris-fretting failures, which is considered as a dominant cause, efforts have been made to work out strict FME program and introduce lower tie plate with debris filters. Regarding a measure to detect failures without delay, an online off-gas monitor is installed, and there occurred no significant secondary failures leading to unnecessary plant shutdowns thanks to the monitor. Current standard fuel designs are 9x9 lattice configuration with the design maximum burnup set at 55 GWd/t and the bundle average discharge burnup at 45 GWd/t. Satisfactory performance of many fuels has gained so far to the burnup level. However recent data show an increase in hydrogen pickup of fuel claddings and spacers in the region of high burnup or high irradiation period, and this is one of the issues that should be addressed to maintain and improve fuel reliability for the reasons that hydrogen concentration may affect ductility of zirconium alloy. Furthermore, a recent study showed abnormally high hydrogen concentration in fuel cladding. Comprehensive root cause analyses were made, which include a poolside measurement for a number of high burnup fuels and development of new technique for tenacious crud removal, with the result that it is confirmed that a unique combination of cladding material, irradiation period and environment caused the phenomenon and it is not expected to occur again. New iron enhanced zirconium alloys, which have been investigated in for a long time, are expected to have greater resistance to hydrogen pickup, and introduction of the alloys is considered as the future materials. Regarding extensive inspections of fuel assemblies and channel boxes at Kashiwazaki-Kariwa NPS, no abnormality due to Niigata-ken Chuetsu-oki earthquake was observed for all plants. This paper deals with the current fuel performance at Japanese

First preliminary design of an experimental fusion reactor

International Nuclear Information System (INIS)

1977-09-01

A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area

International Nuclear Information System (INIS)

Yazid, M; Sutrisno; Sukarman-Aminjoyo; Zaenal-Abidin

1996-01-01

Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area has been done. The aim of this investigation is for tracing the possibility of radioactivity released in the environment during the operation of Kartini reactor. The data was evaluated were monthly monitored data taken from 1986 to 1994 period. The method of analysis was done by comparing the environmental radioactivity data before and after reactor commissioning, off side the reactor up to a radius of 5.000 meters and more than 5.000 meters from Kartini reactor and also compared to the maximum permissible radioactivity according to the current regulation. This evaluation showed that there was no indication of radioactivity release to the environment during this period of reactor operation

Perspectives on reactor safety

Energy Technology Data Exchange (ETDEWEB)

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Perspectives on reactor safety

International Nuclear Information System (INIS)

Haskin, F.E.

1994-03-01

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

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)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

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

Jules Horowitz reactor (RJH): its design

International Nuclear Information System (INIS)

Dupuy, J.P.

2002-01-01

This article presents the design of the new irradiation facility (Jules Horowitz reactor) that is planned to be built on the Cadarache site of Cea. 2 principles have been followed. The first one is based on a physical separation between the systems and activities related to the reactor and the experiments from one hand and the other systems and means dedicated to the treatment of the experimental devices before and after irradiation on the other hand. This first principle implies to build 2 buildings: the reactor building and the nuclear auxiliaries building. Inside the reactor building activities from the reactor itself are separated from those dedicated to experimentation. In order to maximize the efficiency of such a reactor, an important number of simultaneous experiments is expected, which will generate an endless flux of incoming and out-going experiments and as a consequence an important handling work between the different work posts. The second principle aims at easing any handling work without breaking the rules of confinement. The different storing pools, the water pits that lead to the 5 hot cells and the reactor tank will communicate through a water-filled canal that will link the 2 buildings. (A.C.)

Improving the proliferation resistance of research and test reactors

International Nuclear Information System (INIS)

Lewis, R.A.

1978-01-01

Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235 U loading in the reduced-enrichment fuel elements be the same as the 235 U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant

Anaerobic digestion of cheese whey using up-flow anaerobic sludge blanket reactor

Energy Technology Data Exchange (ETDEWEB)

Yan, J.Q.; Lo, K.V.; Liao, P.H.

1989-01-01

Anaerobic treatment of cheese whey using a 17.5-litre up-flow anaerobic sludge blanket reactor was investigated in the laboratory. The reactor was studied over a range of influent concentration from 4.5 to 38.1 g chemical oxygen demand per litre at a constant hydraulic retention time of 5 days. The reactor start-up and the sludge acclimatization were discussed. The reactor performance in terms of methane production, volatile fatty acids conversion, sludge net growth and chemical oxygen demand reduction were also presented in this paper. Over 97% chemical oxygen demand reduction was achieved in this experiment. At the influent concentration of 38.1 g chemical oxygen demand per litre, an instability of the reactor was observed. The results indicated that the up-flow anaerobic sludge blanket reactor process could treat cheese whey effectively.

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

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

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

International Nuclear Information System (INIS)

Dhami, P.S; Yadav, J.S; Agarwal, K.

2017-01-01

Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Kinetic analysis of sub-prompt-critical reactor assemblies

International Nuclear Information System (INIS)

Das, S.

1992-01-01

Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

Nuclear reactor building

International Nuclear Information System (INIS)

Oshima, Nobuaki.

1991-01-01

The secondary container in a nuclear reactor building is made of a transparent structure having a shielding performance such as lead glass, by which the inside of the secondary container can be seen without undergoing radiation exposure. In addition, an operator transportation facility capable of carrying about 5 to 10 operators at one time is disposed, and the side of the facility on the secondary container is constituted with a transparent material such as glass, to provide a structure capable of observing the inside of the secondary container. The ventilation and air conditioning in the operator's transportation facility is in communication with the atmosphere of a not-controlled area. Accordingly, operators at the outside of the reactor building can reach the operator's transportation facility without taking and procedures for entering the controlled area and without undergoing radiation exposure. The inside of the secondary container in the reactor building can be seen from various directions through the transparent structure having the shielding performance. (N.H.)

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

Mirror reactor blankets

International Nuclear Information System (INIS)

Lee, J.D.; Barmore, W.L.; Bender, D.J.; Doggett, J.N.; Galloway, T.R.

1976-01-01

The general requirements of a breeding blanket for a mirror reactor are described. The following areas are discussed: (1) facility layout and blanket maintenance, (2) heat transfer and thermal conversion system, (3) materials, (4) tritium containment and removal, and (5) nuclear performance

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

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

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

International Nuclear Information System (INIS)

Gencheva, Rositsa V.; Stefanova, Antoaneta E.; Groudev, Pavlin P.

2005-01-01

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Status of advanced technologies for CANDU reactors

International Nuclear Information System (INIS)

Lipsett, J.J.

1989-01-01

The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Licensed operating reactors status summary report data as of April 30, 1983. Vol. 7, No. 5

International Nuclear Information System (INIS)

1983-05-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Strengthening IAEA Safeguards for Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-01

During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

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

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

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

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

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

Contributions of research Reactors in science and technology

International Nuclear Information System (INIS)

Butt, N.M.; Bashir, J.

1992-12-01

In the present paper, after defining a research reactor, its basic constituents, types of reactors, their distribution in the world, some typical examples of their uses are given. Particular emphasis in placed on the contribution of PARR-I (Pakistan Research Reactor-I), the 5 MW Swimming Pool Research reactor which first became critical at the Pakistan Institute of Nuclear Science and Technology (PINSTECH) in Dec. 1965 and attained its full power in June 1966. This is still the major research facility at PINSTECH for research and development. (author)

The instrumentation of fast reactor

International Nuclear Information System (INIS)

Endo, Akira

2003-03-01

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

Financial report on the investment and maintenance of the RA reactor and equipment, Annex 5

International Nuclear Information System (INIS)

2002-01-01

For the purpose of maintenance of the RA reactor building and equipment, Serbian Ministry of science, technology and development has approved financial support for the following: purchase and installing 12 new battery cells in the reactor building, for purchase of the missing fire protection equipment, and purchasing the material for protection of the roof of the ventilation center. All the planned tasks were fulfilled [sr

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

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

Safety analysis of reactor's cooling system

International Nuclear Information System (INIS)

1999-01-01

Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

International Nuclear Information System (INIS)

Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

1988-12-01

A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

Action plan during reactor shutdown in October 1965, Annex 5; Prilog br. 5 - Plan radova u toku stajanja reaktora u mesecu oktobru 1965. godine

Energy Technology Data Exchange (ETDEWEB)

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

1965-12-15

The action plan of the division for reactor maintenance during reactor shutdown includes detailed list of tasks for mechanics, electronic and electrical equipment group during the reactor shutdown period in October 1965. It contains tasks for planned shutdown periods in September, August, July, May, April, March, and February 1965. [Serbo-Croat] Plan radova Odelenja odrzavanja reaktora RA za period stajanja reaktora u oktobru mesecu 1965. sadrzi detaljnu listu zadataka masinske grupe, elektro grupe i elektronske grupe. Ovaj prilog sadrzi i zadatke koji ce biti obavljeni tokom planiranih perioda kada je reaktor zaustavljen u septembru, avgustu, julu, junu, maju, aprilu, martu i februaru 1965.

RA Research reactor, Part I: Technical and operational properties of the RA reactor; Analiza sigurnosti rada Reaktora RA I-III, Deo I: Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Zecevic, V; Nikolic, M; Popovic, B; Milosevic, M; Milic, M; Strugar, P; Pesic, M; Nikolic, V; Rajic, M; Radivojevic, J; Jankovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

RA reactor is a research reactor with rather high power density. Apart from research it is used for isotope production and industrial applications due to high reactivity excess (about 11%). It is a thermal reactor, heavy water moderated, cooled by D{sub 2}O, and H{sub 2}O, with a graphite reflector. Nominal power is 6.5 MW. Fuel is 2% enriched metal uranium, reactor core height is 1220 mm, and diameter is 1405 mm. Reactor lattice is square with lattice pitch 130 mm. There is 6 horizontal experimental channels and a graphite column. There is a total of 84 fuel channels and 45 experimental channels in the core. Maximum thermal neutron flux is 5.5 10{sup 13} n/cm{sup 2} s at nominal power level.

Reactor Thermal Hydraulic Numerical Calculation And Modeling

International Nuclear Information System (INIS)

Duong Ngoc Hai; Dang The Ba

2008-01-01

In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

The experimental program of neutronphysics for advanced water reactors

International Nuclear Information System (INIS)

Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

1985-11-01

The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

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

Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

Directory of Open Access Journals (Sweden)

Siniša Šadek

2010-01-01

Full Text Available RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Inherent safety characteristics of innovative reactors

International Nuclear Information System (INIS)

Heil, J.A.

1995-11-01

of the considered innovative designs with passive safety features and increased simplicity, the modular gas-cooled reactors offer the best option for support. 5 figs., 4 tabs., 11 refs., 7 appendices

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community